ML19340B435

Annual Operating Rept of Yankee Atomic Electric Co Nuclear Power Station,1976

ML19340B435
Person / Time
Site:	Yankee Rowe
Issue date:	12/31/1976
From:	YANKEE ATOMIC ELECTRIC CO.
To:
References
NUDOCS 8011100244
Download: ML19340B435 (40)

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O ANNUAL OPERATING REPORT OF YANKEE ATOMIC ELECTRIC COMPANY NUCLEAR POWER STATION ROWE, MASSACHUSETTS 1976 Docket No. 50-29 License No. DPR-3 O

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i TABLE OF CONTENTS

','b Page INTRODUCTION 1

SUMMARY

OF OPERATING EXPERIENCE 1

. CHANGES IN FACILITY DESIGN 3

4 j

CORRECTIVE MAINTENANCE

SUMMARY

- I & C' 7

1 4

CORRECTIVE MAINTENANCE

SUMMARY

- MAINTENANCE 13 LICENSEE EVENT REPORTS 18 I

~ FUEL PERFORMT.NCE 30 t

SUMMARY

OF CONTAINMENT PENETRATION TESTS 31-DATA TABUIATIONS 32 i

j PLANT STATISTICS

-32 i

PLANT GENERATION,RAPH 33 UNIT SHUTDOWNS AND FORCED REDUCTIONS 34-MAN-REM EXPOSURES 35 PRIMARY PLANT. CHEMISTRY 36 LO i

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,e INTRODUCTION.

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L,_j Yankee Rowe is a pressurized water reactor' of '175 MW net dependable

' capacity, owned by Yankee Atomic Electric Company and located in Rowe, Massa chuse tts.

The nuclear steam supply system is a four loop Westinghouse Electric plant. The archit ect/ engineer and constructor for. this plant was Stone & Webster Engineering Corp.

The condenser cooling method is once-through, and the Deerfield River (Sherman Pond).is the ccndenser cooling water source. The plant is subject to license DPR-3 issued July 9,' 1960.

The date of initial reactor criticality was August 19, 1960, and commercial generation of power began July 1, 1961.

There were no major personnel changes during the year at the power station.

SUMMARY

OF OPERATING EXPERIENCE The following is a chronological description of Plant Operations including other pertinent items of interest for the twelve month period ending December 31, 1976.

1-1 At the beginning of the period, the power was limited to 70 percent (120 MW net) pending a full evaluation of the ECCS model by the NRC.

1-5 The upper power limit was raised to 79 percent (140 MW net) by the NRC and the reactor power was increased to that level.

1-13 A three-phase short to ground on the #2 reactor coolant caused an auto reactor trip.

Investigation revealed the failure occurred in

()

the stator windings of this canned rotor pump.

It was determined that Indian Point I utilized a similar pump and with the full coop'ra-tion of Consolidated Edison, one of their pump stators was used as a temporary replacement. The pump was installed and the plant was returned to service on 1-31.

2-2 A pilot wire relay monitoring the Z-126 transmission line operated and opened the line at the IIarriman Station end.

This resulted in a loss of power to the 2400 V bus supplied from the Z-126 line.

The reactor coolant pump supplied from this bus signaled a loss of coolant flow reactor trip. As this was the third such incident of a false signal on the pilot wire, it was opened at the Yankee end to prevent a reoccurrence. Additional redundant relaying still remains to monitor the Z-126 line. The plant was returned to service 'in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and the licensed limit of 79 percent was reached later in the day.

1 2-17 The upper power limit, for ECCS consideration, was raised to 87 percent and the generator load was accordingly increased to about 155 MW net.

4 3-28 The main generator excitor shaf t failed resulting in a loss of excitation trip of the generator concurrent with a reactor scram.

1 1

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A temporcry -replacement exciter was obtaintd from Westinghouse and 4

installed on the generator. The plant was returned to. service on April l' and ' the restricted load limit. of about 155 MW net was achieved by 4-3.

5-6 Outage scheduled to replace the temporary exciter with the new.

repaired exciter. While the unit was shutdown, steam generators number 1'and 4 were opened and tested for tube' leaks. ' The primary-I to-secondary leakage had increased from 3 gal / day to 35 gal / day over a 90 day period. One tube in 'each steam generator showed indication of-leakage and were explosively plugged (refer to LER 76-04). Work

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was completed on schedule,' but startup was delayed 5 days. over NRC licensing matters. The unit _was returned to service on 5-20 and the restricted load limit of about 155 MW net was reached the next day.

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5-24 The generator load was reduced to about 90 Mb for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to permit j

the isolation of. one condenser water box and the plugging of two

-leaking tubes.

6-3 The reactor power restriction was lifted and the reactor power raised to the full licensed level of 600 MW, which corresponds to about 175 MW net electrical.

6-27

- A scheduled load reduction to 130 MW was made for 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to permit surveillance-testing of the turbine throttle and control valves.

i 7 A scheduled load reduction to 130 MW was made for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to permit

- surveillance testing of the turbine throttle and control valves.

7-30 A review of the analysis of the rod ejection accident for Core XI revealed that the methodology used was in error. Upon discovery of i

the error, the reactor power was reduced to 91.5 percent, and the accident analysis was reanalyzed using a correct method. (Refer to LER 76-8) 8-22 A scheduled load reduction to 90 MW was made for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> to permit

. surveillance testing of the turbine throttle and control valves and inspection'of the condenser. water boxes.

9-25

-A scheduled load reduction to 90 MW was made for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to permit surveillance testing of the turbine throttle and control valves and the plugging of one leaking condenser tube, i

10-30 A scheduled outage was made to repair a leak in the #1 feedwater heater. This outage terminated a continuous power run of over 162 days. The plant-was returned to service 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> later on 10-31.

11-28 While coriucting the monthly turbine throttle and control valve-

-surveillance test, the local. operator lost communications with the control room operator. A generator load swing resulted in an auto

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. trip on high moisture separator level. The unit was returned to service-approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> later.

i i-12-28 A scheduled load reduction to 90 MW for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> was made for turbine

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throttle and control. valve surveillance testing.

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CHANGES IN FACILITY DESIGN 7

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1.

The following changes requiring authorization from the Commission were made:

a.

EDCR 74-1, Primary Vent Stack Monitoring System. The change consisted of removing the existing stack monitoring equipment and sealing the PVS penetrations; installing four isokinetic sampling nozzles; locating the particulate, gas, and iodine detector and associated electronics in an enclosure on the PAB roof; and locating remote chart recorders and alarm modules in the main control room.

The change was made to provide monitoring capability for a complete airborne radioactivity sampling and monitoring system.

It providas for continuous off-line monitoring of radioactive particulates, gases, and iodine with the advantage of convenient calibration.

b.

EDCR 75-12, Hot Leg Injection System.

This change consisted of installing a redundant check valve and pipe tee in the vapor container recirculation system between PU-MOV-547, PU-MOV-540 and PU-V-649.

A two inch line containing a manual valve and check was added from the pipe tee to the charging pump suction header on numbers 1 and 2 charging pumps. The change was made to prevent the reactor core boron concentration from reaching a precipitation level and causing possible channel blockage during the long term cooling mode following a LOCA involving a cold leg break.

c.

EDCR 75-28, Supplement 2, Modifications Required to Meet Proposed eN Change No. 117, Supplements 3 (b), 5 and 6.

The change consisted of

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removing No. 1 and 2 auto throw-over switches and replacing them with manual throw-over switches. The change was made to meet the requirements af Proposed Change No, 117, Supplements 3(b), 5 and 6.

d.

PDCR 75-18, Replacement of Rod Position Indication Lamps with LED's.

This change consisted of replacing the indicating light portion of the Primary Rod Position indicating system with light emitting diode (LED) style indication per Proposed Change 134.

The change was made to improve the reliability of the system such that the position of the rods is indicated within the desired accuracy of

+3 inches.

e.

PDCR 75-21, Addition of a Pulsation Dampener on the Number 3 Charging Pump Discharge Header. The change consisted of the addition of a surge dampener and the replacing of corresponding pipe components with large sweeping clbows and lateral in the discharge header of No. 3 charging pump. This change was made to reduce the intensity of the vibrations experienced by the piping, in order to reduce the possibility of weld failures.

2.

The following changes not requiring Commission approval were made:

a.

PDCR 73-10, Installation of VC Penetration Test Taps. The change consisted of the installation of manual valves and test taps in the Continuous Leak Monitoring, M.C. Vent and Valve Stem Leak-off Systems.

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This change was made to allow duplication of accident conditions for performing penetration tests in accordance with Appendix C of the Plant Technical Specifications.

3

b.

PDCR 73-16, Change to Main Coolant Cbeck Valve Internals. The

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(3 wear points of the spare clapper assemblies for the main coolant

(_)

check valves were replaced with wear resistant ones.

This change was made to provide longer life for the internals and subsequently decrease the maintenance frequency for their replacements.

c.

PDCR 73-19, Pressurizer Heise Line SI Initiation Pressure Switch and Excess Flow Check Valve Relocation. The change consisted of moving the Safety Injection Initiation Pressure switch from the valve room of the Primary Auxiliary Building to a location outside the Biological Shield at the broadway level on the pressurizer compartment missile shield. A limited flow check valve was added downstream of the sensing line tee of the Safety Injection Initiation Pressure Switch on the Heise Gage sensing line which was also re-arranged. The changes were made to minimize the time between the

?ressurizer Pressure reaching 1700 psi and the Safety Injection Initiation Pressure Switch sensing 1700 psi.

The limited flow check valve was added to the Heise gage line so that a break in the line between the Vapor Container and the upper level PAB would not result in an uncontrolled leak.

d.

PDCR 73-20, Addition of Steam Generator Blowdown Isolation Valves.

The change consisted of adding two inch, globe, isolation valves in the blowdown lines near the steam generators.

To allow the installation of this valve as close as possible to the steam generator, the one inch drain line was cut out and capped at the blowdown line tee and steam gener ator. The change was made to allow servicing the blowdown line without draining the steam generator.

-s qj e.

PDCR 74-2, Secondary Calorimetric Feedwater Flow Instrumentation.

The change consisted of installing four new flow indicating and totalizing channels,in parallel with the existing Bailey flow de tec tors. This change was made to increase in accuracy the plant calorimetric determination.

f.

PDCR 74-6, Repair of Charging Line Relief Valve SV-209.

The change was accomplished in two steps. First, the leaking relief valve was replaced with a spare valve that had had a new nozzle and disc installed and had a flange welded on its inlet connection. At a later date, the discharge was fitted with flanges.

The change was made to facilitate repairs and reduce down time.

g.

PDCR 74-12, Instrumentation valve Add itions. The change consisted of adding second isolation valves and/or test valves to various instrumentation in the plant.

This change was made so the instrument calibration could be conducted without requiring a complete plant cooldown and also decrease the time needed for calibration.

h.

PDCR 74-14, Installation of Indicating Fuses and Removal of Auto-Tie Function on #1 and #2 Battery Charger Circuits. This change consisted of the addition of a small auxiliary fuse in parallel with the main DC fuses that would trigger a microswitch which in turn would initiate an alarm when the main fuse blows. The automatic r

closing feature of the charger's DC breaker was removed.

This change l (s)/

was made to reduce the possibility of accidentally discharging l

Nos. 1 and 2 station batteries.

4 L

i. PDCR 75-5, Steam Generator Narrow Range Level Trip System Change.

The change modified the Sigma Indicator / Alarm units, on the Steam t

Generator Narrow Range Level Trip System, by removing the incandescent lamps and associated. photocells and installing light emitting diodes.

and a new type photocell. A second Sigma Indicator unit (non-controlling) was mounted on panel 7F of the MCB.

The change made it possible to remove the Indicator / Alarm unit for maintenance and will reduce the failure frequency of the light indication.

j. PDCR 75-8, Drain Valve Addition to the Chargingg Line Between CH-MOV-523 and CH-V-611.

The change consisted of the addition of a 3/4" valve on the bottom of the 21/2" charging line between CH-MOV-523 and CH-V-611.

This change was made to allow draining of the charging line betwec n CH-V-611 and CH-MOV-523 for maintenance.

i k.

PDCR 75-12, Fan Room Waste Gas Lines from OH Condenser and Decay Filter Change. The change consisted of removing the isolation value from the OH condenser, isolation valve from the Gas Decay Filter, and associated piping to FN-ll.

The two lines were repiped into a common line via VD-V-982 and SOV-310 to terminate in a new 17" duct installed as part of EDCR 74-3.

This change was made to facilitate the completion of EDCR 74-3 because the existing piping was in the way.

1.

PDCR 75-13, Electrical Connector Addition in the Diesel Fuel Oil Level Control System. An electrical connector was added to the day tank level control system between the level switches and the junction box for each emergency diesel generator fuel oil day-tank.

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The change allows disconnecting the level switches for calibration and surveillance without disconnecting the wires in the junction box.

m.

PDCR 75-14, Addition of an Isolation Valve and Test Connection to the Primary Seal Tank Pressure Sensing Line. The change consisted of i

adding an isolation valve and test connection with associated valve to the Primary Pumps Seal Water Tank pressure sensing line. The change was made to allow calibration of PI-234 and PT-236 without requiring the isolation of the makeup water connection to the seal water tank.

n.

PDCR 75-15, Boric Acid Mix Tank Remote Level Indication. The change consisted of using the spare indicator in the main control board for remote level indication of the Boric Acid Mix Tank (BAMT). The indicator tubing of the old Safety Injection Pump Discharge Pressure was used for transmission of the signal. This change was made to provide a means for the control room operator to evaluate the condition of the BAMT, thus providing for a safer and more reliable operation of the Chemical Shutdown System and the plant.

o.

PDCR 75-17, Installation of a valve at the Cavity Recovery Connection in the Main Coolant Drain Header. The change consisted of installing a 3/4 inch angled cap valve, with a capped nipple, to the previously capped cavity recovery _ connection of the main coolant drain header.

The change was made to facilitate hydrostatic testing of inter-

' C, 3

,/

connected systems thereby reducing the overall maintenanca time and decreasing the cost of the performed maintenance.

5

p.

PDCR 75-19, Replacement of MCB Information Lights. This change

~g consisted of replacing the safety injection pump and valve status

(_)

lights system on the nuclear section of the Main Coolant Board with an equivalent but mare reliable type system. This change also involved replacing.the dropping resistors and properly mounting the new ones. The change was made to increase reliability and case of maintenance, thus, decreasing overall maintenance time 2nd cost.

q.

PDCR 75-23, Installation of Locking Device on No.1 or 3 Charging Pump Speed Control Mechanism. This change consisted of installing a locking device capability on Nos. 1 and 3 charging pumps' speed controller. This change was made to ensure that the pump is not accidently placed back in the variable speed mode.

Technical Specifications, D.2.e.5, requires that two fixed speed pumps be available at all times when main coolant pressure is 11000 psig.

PDCR 76-9, Installation of Thermocouples in the Boric Acid Mix Tank r.

(BAMT) Suction Line. The change consisted of installing a temperature monitoring system for the line from the BAMT to the charging system suction line. The system included twelve Iron /Constantan Type J Thermocouples, a thermocouple selector switch and a thermocouple indication ' me ter.

This change was made to permit operation'of i

the plant in. compliance with the Technical Specifications.

s.

PDCR 76-14, Modifications to the Incore Instrumentaticn (ICI) System.

The change consisted of installing a pressure switch on the CO2 supply to the IC Flux Mapping System transfer device frame and an associated alarm light on the IC Flux Mapping System cabinet in the

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control room, as well as, an isolation valve and associated tubing on the leak alarm system of the IC Flux Mapping System. These changes were made to give a warning of low CO2 pressure so the system will noc be operated without the necessary moisture inhibiting gas and to save time in performing the surveillance of the Leak Alarm.

t.

PDCR 76-23, Replacement of the Boric Acid Mix Tank (BAMT) Heat Trace.

The change consisted of replacing the Type GE silicone 19 AWG Si 53921 with two, separately powered, circuits of Chemflex. This change was made to provide a more reliable means of maintaining the Boric Acid solution, in the pipe under the BAMT, at a temperature 1150 F as required in the Technical Specifications (3.1. 2. 3).

u.

EDCR 75-5, New Fuel Storage Rack Modification. The change modified some of the new fuel storage racks by adding an aluminum channel lined' with rubber to the rack. The change was made to provide support for the new Exxon fuel over its entire length.

v.

EDCR 75-10, VC Purge System Modification, was completed during the report period. The change consisted of installing a 15,000 cfm booster fan and associated duct into the inlet duct of the existing vapor container ventilation and purge system.

The electrical work l

done consisted of installing the fan controller on the motor enclosure, running conduit and cable, and installing a start /stop switch' in the control room. This change was made to provide better control of the supply and purge air flow in order to reduce the danger s_/

of 'both inadvertent-ground level radioactive releases and physical-danger to personnel.

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e CORRECTIVE MAINTENANCE

SUMMARY

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I&C Corrective Action System Component

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Malfunction Effect on Taken to Prevent Cause Result Safe Operation Repetition Nuclear Channel # 8 Loose dial Could not adjust Instrument. Fine Gain gain Control Main Steam PCV-402 Deteriorated Leaked by seat and disc.

Nuclear Channel #6 Metal shavings Meter stuck Instrument. MCB meter in meter move-ment Steam

1. 2 S/G NR Out of Calib.

Inaccurate Generating indication Safety

1. 1 SI Flow Deteriorated Leaked Injection Transmitter fitting Safety

1. 1 SI Flow Deteriorated Leaked Injection Transmitter fitting

,_ Safety

1. 1 SI Flow Deteriorated Leaked I\\.

Jnjection Transmitter fitting Prima ry T-ave Meter Defective Incorrect meter operation Nuclear Meter & test Ground connec-Ground on Vital Instrument. Panels tion on test Bus and blip pulse generator on MC Pressure low Primary Ion Exchange Deteriorated Leaked by Flow D/P Cell seat Primary Seal Tank Defective Incorrect Pressure Meter Indication Indication Reactor

1. 1 S/G NR Deteriorated High Alarm did Protection Level Trip components not reset System Safety SI Tank Calibration Incorrect Injection Pneumatic Drift indication Level Channel

<-irimary PI-230 Deteriorated Leak

(_)

fitting Panalarm Power Supply Deteriorated Incorrect

1. 28114 components output voltage 1

7

- CORRECTIVE MAINTENANCE

SUMMARY

OQ I&C (continued)

Corrective Action Malfunction Effect on Taken to Prevent System Component Cause Result Safe Operation Repetition Panalarm Power Supply Deteriorated Incorrect

1. 23115 components output voltage Nuclear Nuclear Deteriorated No chart Instrument. Recorder bearing on the advance Chart Drive Motor Safety

1. 1 Loop.

Insufficiently Mirute leak Injection Pressure compressed Transmitter fitting Safe ty Spare Flow Defective Incorrect Injection Transmitter Component Operation Reactor

1. 3 S/G NR Level Deteriorated High alarm Protection Trip Photocell would not 7

reset (bclear Spare recorder Deteriora ted Improper operation Instrument. amplifier tubes Feedwater & Spare amplifier Deteriorated Improper operation Condensate

1. 001 tubes Safety SV'SOV #231744 Deteriorated Improper operation Injection coil 4

Feedwater & Spare Bailey Deteriorated Improper operation Condensate receiver #24 Components Feedwater & Spare Amplifier Deteriorated Improper operation

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Condensate

1. 003 tubes Feedwater & Spare Bailey Condensate Amplifier #002 Feedwater & #3 Feedwater Loose breech Leak Condensate regulating valve Nuclear Channel #8 Loose set Knob loose Recorder screw on shaft

' g-9harging Vari-orifice Calibration-Vari-orifice (m/

Drift did not close sufficiently J

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.s CORRECTIVE MAINTENANCE SUntARY O-I&C (continued)

Corrective Action Malfunction F;ffect on Taken to Prevent System Component Cause Result Safe Operation Repetition Main Steam PCV-402 Calibration Ieaked by Drift i

Nuclear Nuclear Cover Loose Intermittent l

Instrument. ' Recorder operation Primary Spare level Deteriorated Incorrect amplifier compensation operation switch Primary Spare pressure Deteriorated Incorrect amplifier high adjustment operation pot Reactor

1. 2 S/G NR Defective alarm High level Protection Indication /

pickup alarm point Alarm incorrect Reactor

1. 2 S/G NR Alarm unit out Alarm would Protection Level Trip of adjustment not clear actor

1. 3 S/G NR Unkr ~

Alarm would Protection Level Trip not clear Nuclear Spare C Defective Incorrect Ins trument. Amplifier tube operation Reactor

1. 1 S/G NR Trip Defective High High alarm Protection Indicator /

& Low Alarm would not Alarm Units cancel l

Safety SI Tank Static on Erratic level l

Injection Invel meter face indication indication Primary Charging Deteriorated Leaked flow ball in vent l

Transmitter valve Feedwater & #3 Feedwater Marks on Leaked Condensate Regulating packing Valve seat Safety SI Tank Defective Grounded Injection Level Terminal Board

, O clear Spare Pulse Weak tubes Noise

' NIstrument.

Int. Drawer sensitive 9

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CORRECTIVE MAINTENANCE

SUMMARY

,(j I& C (continued)

Corrective Action Malfunction Effect on Taken to Prevent Syntem Component Cause Result Safe Operation Repetition Nuclear Spare Prop.

Weak tubes Noise Instrument. Counter Power sensitive Supply Drawer Nuclear Spare Log. MIC Weak tubes No_ne Instrument. Drawer sensitive Nuclear Spare Scram Weak tubes Burns out Instrument.

Amp.

bulbs Primary Spare Tave Two glass Meter sticks Instrument. Meter plates in-stalled instead of one Primary MC Press. Low Deteriorated Grounded Instrument.

Press Alarm coil Alarm Reactor

1. 3 S/G NR Alarm pickup Continuous I Irotection Level Trip failure high level

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System alann Service PS-414 Deteriorated Minute Water fitting leak Panalarm Power Supply Deteriorated Incorrect

1. 28113 components output Safety SI Tank Level Deteriorated

'41nute Injection Indication fitting leak Lower PAB Purifica-Purification Deteriorated Minute tion Pump Discharge fitting Leak Press. Ind.

Pressurizer Presa. Ind.

Deteriorated Minute on Heise Line fitting Leak Safety PI-NS-4 Deteriorated Leaked by Injection valve Nuclear Spare UIC/CIC Defective Incorrect Instrument. Enclosure Detector and operation n

cables

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'a Feedwater CV-1000,1100 Loose packing Leak 1200 and 1300 10

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CORRECTIVE MAINTENANCE

SUMMARY

I & C' j

(continued) l

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Corrective Action i

Malfunction Effect on.

Taken to Prevent l

System Component

~ Cause-Result

-Safe Operatior.

Repetition 4

i Waste PI-318 Deterorated Minute Disposal fitting leak Purifica-DPI-202 Deteriorated Bonnet tion valves Leak l

1 Was te PRC-302 Calibration Erratic Disposal Drift Operation Radiation Gamma Defective HV Unit Monitor Guard Control Drift J

j l

Safety Loop #2 Calibration Indicated S

Injection Flow Drift High Safety llP SI Static Read %250 Injection Flow gpm Pressurizer Pressure Weak Indiciated I

Recorder Battery high ste PRCV-102 Reset Drift Sluggish Disposal Operation i

Nuclear Nuclear Broken Pen No Printout 1 Instrument. Recorder 1i Feed and

1. 2 S/G NR Deteriorated Improper Condensate Compensating Booster.

operation i

Relay

Auxiliary PCV-604 Dirty Position-Improper Steam ner Pilot Valve operation 4

Primary.

Spare Tempera-Deteriorated Improper ture Amplifier Potentiometer operation and broken wires

.i

'I Pressure Pressurizer Tubing Vibration Rubbed side

. Installed rubber Control Level (NR) Line l,

gasket around line i Bailey Spare Booster Relays Defective-Incorrect j

components operation Waste.

~PRC-303

' Incorrect Erratic "Asposal Proportional operation Band Setpoint Feed and

1. 2.Feedwater Deteriorated Minute Condensate Regulating Breech Block leak t

Valve 11

e CORRECTIVE MAINTENANCE

SUMMARY

,m I&C (continu ed)

Corrective Action Malfunction Effect on Taken to Prevent System Component Cause Result Safe Operation Repetition Safety SI-PR-58 Setpoint Drift Maintained Injection Low Pressure Feed and Bailey Deteriorated Improper Condensate Booster #BB-5 Component operation Nuclear Log Micro-Defective tubes Incorrect Ins trumen t, ammeter Drawer operation

1. 5Y Radiation VC APD Deteriorated Solenoid Monitor Solenoid Seating Leaked by surfaces Nuclear IR #4 Log Deteriorated Indication Instrument. Microammeter Log Diode or Drift Electrometer tube Safety SI-PR-58 Setpoint High Accumula-( njection SI-PR-59 Drift tor Pressure Alarm Purification PU-FICV-202 Calibration Incorrect Drift Controller

Response

Feed and

1. 2 Bailey Erroded Breech Block Condensate Control Valve Breech Block Leak Instrument Air lines to Too small Slow Installed larger Air TV-401A,B,C Operation line

&D Emergency DG Day Tank Incorrect Controls Installed bush-Power Iovel position level too ings to raise Controllers low level controls 42" Safety LT & LI-401 Calibration Incorrect Injection Drift readings Compressed SOV-00639 Deteriorated Leaked by Air Internals f c fety SI Tank Level Partially Incorrect Insulated exposed a

,(_,hjection Transmitter exposed sensing level section of line line indication 12

)

CORRECTIVE MAINTENANCE

SUMMARY

Akh MAINTENANCE Corrective Action Malfunction Effect on Taken to Prevent System Component Caune Result Safe Operation Repetition Vital Bus Inverter Worn bruches Primary

1. 2 Main Galled Turns hard Coolant Pump Vent Valve Safety SI-V-599&600 Insufficiently Minute leak Injection compressed packing Control Rod Control Rod Deteriorated Rods would Drive Drive Contactor not move in or Circuitry out manually Safety CS-y-621 Insufficiently Leaked Injection torquod bonnet Charging

1. 1 Charging Deteriorated Leaked Pump cylinder cover gasket i-,,

(_,% Recirc.

PU-MOV-544 Loose Contacts Would not close electrically i

Pressurizer HSS-20A Loose oil Leaked reservoir nut Charging CH-V-611 Deteriorated Leaked bonnet"O" Ring Vital Bus Frequency Dirty Rheostat Meter Meter Oscillations Main Coolant MC-MOV-503 Deteriora ted Excessive stem packing leakoff Main Coolant MC-MOV-505 Galled union Difficult to turn Charging

1. 3 Charging Cracked welds Relief valve Pump line leaked Prescurizer PR-MOV-512 Deteriorated Leaked by Seating surfaces

(~"ppor Inner Deteriorated Leaked

(,stntainer Personnel gasket Hatch 13 1

1

CORRECTIVE MAINTENANCE

SUMMARY

MAINTENANCE (continued)

Corrective Action Malfunction Ef fect on Taken to Prevent System Component Cause

, Result Safe Operation Repetition Charging

1. 3 Charging Cracked weld Leaked Pump Charging

1. 1 Charging Normal Excessive Pump deterioration gland leakage of packing Cha rging

1. 2 Charging Normal Excessive Pump Deterioration gland leakage of packing Charging

1. 3 Charging Normal Excessive Pump deterioration gland leakage of packing Safety CS-V-668 Deteriorated Stem Injection packing leakage Safety CS-MOV-535 Deteriorated Stem leakage f Injection packing UCharging CH-V-620 Bad weld Leaks Charging

1. 1 Charging Deteriorated Leaks Pump Relief gasket, flange Viv and bellows assembly Charging Spare Charging Defective Unavailable as Pump helief nozzle a replacement Vlv Main Steam Drain line Deteriorated Pin hole pipe leak Pressurizer PR-V-606A Deteriorated Leak packing Feedwater & BF-V-650 Deterioration of Stem leakage Condensate packing Main Coolant VD-V-712 Deterioration Steam leakage of packing Main Coolant VD-V-724 Deterioration Stem leakage

(~S of packing O

14

CORRECTIVE MAINTENANCE SUMtARY MAINTENANCE (continued)

Corrective Action Malfunction Effect on Taken to Prevent System Component Cause Result Safe Operation Repetition Charging

1. 1' Charging Deteriorated Cap leaked Pump Suction gasket J

Strainer l

Auxiliary TV-405

' Deteriorated Leaked by l

Steam threads Charging

1. 1 Charging Deteriorated Valves making I

Pump valves excessive noise Main-MC-V-310 &

Packin~

Hot stem leak

.i Coolant CS-V-538 adjustment off Diesel

1. 1 D/G Clamp Hose leak l

Generation adjustment 4

Safety NS-V-14 Insufficiently Leaked Injection compressed packing harging

1. 1 Charging Loose bolts Pump head and Pump strainer cap leaked Safuty SI-V-621 Loose bolts Bonnet leak Injection Feedwater Emergency Insufficiently Excessive Boiler Feed compressed leakage i

Pump packing from middle ram Main MC Pump Surface defects Coolant Blank Flange 1

Exciter Exciter Broken shaft Commutator damage Main Steam HP Steam Trap Defective Leaked i

Line in Orifice wold House Main MC-MOV-501 Packing Leaked i

MC-MOV-532 con 91etely corpressed t

e i ge"afety CS-MOV-539 Seat backed Would not

! (,[hje ction out close completely 15

.~-

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

I j

. CORRECTIVE MAINTENANCE

SUMMARY

MAINTENANCE' (continued)

Corrective Action Malfunction Effect on Taken to Prevent l

System Component Cause Result Safe Operation Repetition Sample SA-V-513 Deteriorated Leaked by Internals i

Primary PR-V-607 Deteriorated Bonne t gasket leak i

Control Rod BK-2 Unknown ABC would Drive not close i

Charging

1. 3 Charging Cracked weld Leaked Pump Relief r

Valve Safety

1. 1 LPSI Deteriorated flot packing j

Injection-Pump packing gland s

VC Recirc.

Piping Insufficiently Minute leak Compressed i

Gasket t

' QrCharging CH-V-690 In::ufficiently Minute leak Compressed Packing f

Safety Flow Orifice Loose Flange Minute leak i

Injection in CS-MOV-634 Bolts Line Surface defects Main MC Pump Coolant Blank Plange Charging Discharge Fatigue Weld leak Added a surge header No. 3 dampener and i

j charging pump sweeping elbows and lateral per PDCR'75-21 l

Purification #2 Purifica-Inose plug Minute leak tion pump

. Charging

1. 3 Charging Deteriorated

' Leak Pump weld l

Purification #2 Purifica-Clogged Excessive seal

. tion pump cooling line temperature Main Steam Right hand Deteriorated Pin hole throttle body' leak valve drain trap 16

CORRECTIVE MAINTENANCE

SUMMARY

MAINTENANCE

(,)

(continued)

Corrective Action Malf unction Effect On Taken to Prevent System Component Cause Result Safe Operation Repetition Main No. 3 Main Insufficiently Minute packing Coolant Coolant compressed leak pump vent packing valve

!!ydrogen Post Accident Loose Possible Control 11 Control Fans connection reduced air 2

duct flow Ilydrogen No. 1 VC Loose louvres Fan rotated Control

!! Vent Fan backwards 2

Charging

1. 3 charging Deteriorated Reduced pump internal valves capacity and springs Main MC-MOV-501 Insufficiently Packing leak Coolant compressed packing

[, Feed &

1. 2 Feedwater Unknown IIanger &

Oondensate line pipe support shif ted hanger Emergency

1. 1 Diesel Insufficiently

linute leak Power Generator tight radiator hose clamp Emergency

1. 2 Diesel Small crack Slight leak Power Generator in exhaust pipe

.f g

L) 17

=

. _ ~ _ =..

1 i

LICENSEE EVENT REPORTS

O The following is a chronological listing of all the LER's submitted during the report period.

{

1. 76-1 DESCRIPTION OF OCCURRENCE l

During normal plant operations the auxiliary operator noted slight leakage from the No. 3~ charging piping in two areas:

Relief valve discharge 4

flange to piping weld, and pump suction and pump suction tee to the relief valve discharge line welds.

1 Charging was transferred to the No. 1 charging pump and the No. 3 charging pump was shut down and isolated.

This event was reported in accordance with Appendix A, Technical Specification Section E.2.b.2.

Conditions leading to operation in a degraded mode permitted by a limiting condition for operation.

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE 1

The failure of the welds did not impose an unreviewed safety question or hazard to the health and safety of the public in that the leakage was minimal and contained in the Cubicle Area of the Primary Auxiliary Building which is exhausted to the Primary Vent Stack. Safe operation of the plant l()

was not affected in that two (2) other charging pumps were available for service.

CORRECTIVE ACTION:

The leak on the relief valve flange was corrected by installing a new flange and nipple. The leak on the suction line was corrected by installing a new tee and fittings, i

There have been previous failures on this system that can be attributed to vibration and PDCR 75-21 has been initiated to correct the vibration.

~4 FAILURE DATA:

Previous failures and malfunctions on the charging systems that can be attributed to vibration induced failures are numerous, Ref. A0 75-07.

1. 76-2 DESCRIPTION OF OCCURRENCE

{

During normal plant operations the Auxiliary Operator noted a leak in the weld.on a 2-1/2 in to 1-1/2 in, reducer on the discharge header of the No. 3 charging pump.

Charging was transferred to the No. 1 charging 4

pump and the No. 3 charging pump was shut down and isolated.

18 1

-e

p

()

This event was reported in accordance with Appendix A, Technical Specifica-tion, Section E.2.b.2, conditions leading to operation in a degraded mode permitted by a limiting condition for operation.

DESIGt!ATIO!! OF APPAREllT CAUSE OF OCCURRENCl; The apparent cause of the crack in the reducer's weld was vibration induced fatigue failure caused by continuous operation of the positive displacement charging pump.

ANALYSIS OF OCCURRENCE The failure of the weld did not impose an unreviewed safety question or hazard to the health and safety of the public in that the leakage was minimal and contained in the cubicle Area of the Primary Auxiliary Building which is exhausted via the Primary Vent Stack. Safe operation of the plant was not affected in that two (2) other charging pumps were available for service.

CORRECTIVE ACTION The crack was on the 1-1/2 in, side of a 2-1/2 in. to 1-1/2 in, reducer.

The defect was Liquid Penetrate Examined and ground out.

The area of the defect was rewelded and the repair was then radiographed and hydro tested satisfactorily prior to return to service.

There have been previous failures of this system that can be attributed to I

vibrations. PDCR 75-21, Addition of a Pulsation Dampener on the No. 3

\\)

Charging Pump Header which was initiated to minimize the effects of the vibrations has been approved by the Manager of Operations and is now awaiting NRC approval.

FAILURE DATA Previous failures and malfunctions on the charging system that can be attributed to vibration induced failures are numerous, Ref. RO 76-1.

1. 76-3 DESCRIPTION OF OCCURRELCE While shutdown for the main generator exciter repair, coincident review of the ECCS analysis revealed that the flow capability of the LPSI accumulator was somewhat greater than that assumed in the LOCA analysis.

This greater flow was assumed to be conservative, though, it was in fact less conservative.

This event was reported in accordance with Technical Specification E.2.a. (7).

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE The apparent cause of this event was an error in the assumption of the accident analysis.

A 19

ANALYSIS OF OCCURRENCE Appendix K criteria requires that credit cannot be taken for the ECCS coolant discharged during the bypass phase of the transient. Thus with a greater accumulator flow rate, a larger fraction of its inventory must be disregarded, and the accumulator empties sooner. A significant reduction in this accumulator inventory would yield a peak clad temperature beyond the limits of 10 CFR 50.46; thus, the reactor could have been operated in a manner less conservative than assumed in the analysis.

CORRECTIVE ACTION a

i A reanalysis was performed to correct the discrepancy. Following this, a proposed change to the Technical Specifications was submitted to reduce the ECCS accumulator pressure and, thereby, reduce the flow to be consistent with that assumed on the accident analysis. The proposed change was approved and the Safety Injection accumulator pressure was changed from 410 psig to 337 psig. The reactor was then returned to power operation.

FAILURE DATA No record of previous failure. See Proposed Change 138 and supplements for a complete description of this event.

1. 76-4

[

DESCRIPTION OF OCCURRENCE During the course of normal operation, the primary to secondary leakage from No. 's 1 and 4 steam generators was noted to increase from %3 gal / day to N33 gal / day over a 90 day period.

During a scheduled outage, the associated loops were drained, manways opened and the secondary side hydrostatically tested to check for leaks.

The test revealed one tube in each steam generator to be leaking:

No. 1 S/G Tube K-22 No. 4 S/G Tube M-44 This event was reported in accordance with Technical Specification E.2.a. (3).

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE The apparent cause of this event is suspected to have resulted from the use of phosphate chemistry control in the early years of operation.

The degradation of the 304 S.S. Steam Generator tubes has been noted to be a definite problem area between the tube sheet and 4 inches above the tube sheet.

')

(J 20

1 l

i 4

l ANALYSIS OF OCCURRENCE v

the consequence of the increased leakage would have been an increased l

release of radioactivity.to the environment and an increased depletion l

rate of the ventilation filters.

CORRECTIVE ACTION The two leaking tubes were explosively plugged, on the inlet and outlet side, by Westinghouse personnel.

FAILURE DATA The degradation of the steam generator tubes has been noted over the past 9 years of power operation. The Steam Generator tubes plugged to date are shown on Table 1.

1. 76-5 j

DESCRIPTION OF OCCURRENCE At approximately 0700 during a routine startup the auxiliary operator noted slight leakage from the weld on the inlet to the No. 3 charging pump relief valve.

J Charging was transferred to the No. 1 charging pump and the No. 3 charging i

pump was shut down and isolated.

This event was reported as Reportable Occurrence 50-29/76-05 in accordance

]

with Technical Specifications E.2.b. (2).

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE i

i The weld failed in the joint between a 1-1/2" x 1" 6000# reducing insert and a section of 1" Schedule 80 stainless pipe.

It appeared that a full 1/16" pullback gap was not allowed during the installation of the pipe.

The resultant stress rise across the weld caused the weld to fail upon experiencing the induced vibrations of the fluid during full speed pump operation.

1 ANALYSIS OF OCCURRENCE 1

I The failure of the welds did not impose an unreviewed safety question or hazard to the health and safety of the public in that the leakage was minimal and contained in the Cubicle Area of the Primary Auxiliary i

Building which is exhausted to the Primary vent Stack. Safe operation i

of the plant was not affected in that two (2) other charging pumps were i

available for service.

CORRECTIVE ACTION The'affected section of pipe was cut out and replaced. The plant ensured that the 1/16" minimum pull back gap was maintained.

21 1

._~

~

1 I

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t i

FAILURE DATA i

! [~

. There are not previous records of this type o'f failure, however, several l

welds have failed due to' excessive vibrations as reported in RO 76-1 and 75-07.

j.

1. 76-6 l

DESCRIPTION OF OCCURRENCE j

During normal operation a routine safety injection tank makeup was performed j

using approved plant procedures. The supervisor completing procedure OP-2654,

" Safety Injection Tank Makeup", contacted higher supervision indicating j

that he could not meet the boron concentration of the acceptance' criteria (200 + 50 ppm). The baron concentration was above the acceptance criteria.

The concentration could not'be decreased because of recirculation time limitations imposed by Technical Specification change No. 126.

i

. DESIGNATION OF PROBABLE CAUSE OF OCCURRENCE i

The event was caused by personnel error. Prior to Core XII, the only i

boron limit for the safety injection tank was a 2200 ppm minimum concentration.

Supplement 3 to Change No. 117 added a' maximum boron concentration of 2400 l

ppin.

The administrative limite was 2300 + 50 ppm.

l Plant personnel in reviewing the completed procedures, did not note that the administrative limit and acceptance criteria had been exceeded.

l ANALYSIS OF OCCURRENCE The consequence or potential consequences from the standpoint of public j

i~

health and safety are minimal.

Recent re-analysis of the maximum allowable boron concentration limits shows that the concentrations experienced were well within the safety analysis limits.

CORRECTIVE ACTION All personnel involved in the performance and review of procedures have j

been~ reminded of the necessity of careful and accurate procedural compliance.

The safety. injection tank boron concentration was lowered by dilution and mixing on July 1, 1976.

FAILURE DATA l

Safety injection tank makeups resulting in concentrations out of the Acceptance Criteria band occurred on the following dates:

Date Concentration i

• 12/13/75 2369 12/28/75 2353 5

1/30/76 2377 2/13/76 2372 3/20/76 2377' 4/28/76 2427 5/8/76 2411 1

(

5/14/76 2400 6/11/76 2408 4

(

6/24/76 2400-1

• Plant in a shutdown condition.

22

~

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(_s)

1. 76-7 DESCRIPTION OF OCCURRENCE During normal operation, an investigation was initiated into the details of Peportable Occurrence 76-06, Personnel Error.

The investigation discovered that during the hydrostatic leak test of the reactor vessel and portions of the Physics Testing program for Core XII startup, the assumed boron concentration of the LOCA Analysis was exceeded.

Amendment No. 21, Enclosure A to the Technical Specifications assumed maximum boron concentration limits for the main coolant, safety injection tank and safety injection accumulator to minimize the effects of boron precipitation following a loss of coolant accident. With the boron concentrations in excess of these assumed on the analysis, the post LOCA sump and core boron concentration would have exceeded that assumed in the analysis.

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE The apparent causes of this event are administrative and personnel error in that the main coolant boron concentration limit for Core XII was not clearly stated as a maximum limit.

Personnel, in reviewing the Proposed Change, did not perceive or grasr the full significance of these limits (assumptions), as given.

n

(]

ANALYSIS OF OCCURRENCE The review of this event showed that the highest boron concentration was 2420 ppm and occurred during the main coolant system hydrostatic test.

The evaluation and bases for the boron concentration limit is given in the memo from W. J. Szymczak to J.

L. French dated July 19, 1976 (File SAG 76-152) entitled " Boron Concentration Limits for Yankee Rowe" and shows that the boron concentration of 2413 ppm was based on a volume weighted average value of main coolant, refueling water storage tank, and accumulator concentrations. Furthermore, the analysis was based on the reactor being at 100% power (600 MWt) for an infinite time and assumed the worst case which is during the latter part of core life. The evaluation further shows that at the beginning of core life maximum allowable weighted boron concentration was 11,250 ppm.

Therefore, based upon the above discussion the consequence of this event from the standpoint of public health and safety is minimal in that the boron concentration was within an acceptable operational limit although above the assumed limit of Proposed Change 117, 4

Supplement 3, Enclosure A, (Amendment No. 21).

CORRECTIVE ACTION As the boron concentration presently is in compliance, and has been since startup of power operations for Core XII, no corrective action could be taken. To prevent reoccurrence, the responsible persons have been reminded of the importance of recognizing accident analysis assumptions and

(~S translating them into plant operating limits.

'm..)

23

<~3 FAILURE DATA See RO 76-6, Unacceptable ST Tank Ik>ron concentration.

N76-8 DESCRIPTION OF OCCURRENCE A review of the analysis of the rod ejection accident for Core XI revealed that the methodology used was in error.

The analysis of the rod ejection accident for Core XII was based on the Core XI analysis so it too was not properly calculated. Upon discovery of the error, the accident analysis was reanalyzed using a correct method.

DESIGNATION OF APPALENT CAUSE OF OCCURRE!!CE Personnel error in selection of the methodology for use in the accident analysis caused this event.

ANALYSIS OF OCCURRENCE The results of the reanalysis indicate that the maxinum allowable zero power ejected rod worth is 0.75% p.

'1he original rod worth assumed was 1.01 p, hence the analysis for zero power control rod ejection accident was non-conservative. The full power rod ejection analysis was conservative using the original method.

Both the actual predicted and measured ejected

~s k.)

rod worths were less than 0.75% p, therefore, Core XII as compared to the reanalysis.

CORPECTIVE ACTION Operation of the plant was verified to be within the new analysis results.

A proposed change to the Technical Specifications (#125, Supplement 9) was drafted to correct the license.

FAILURE DATA There hav., been no previous errors in the rod ejection portion of the accident analysis.

1. 76-9 DESCRIPTION OF OCCURRENCE The Yankee Rowe Core XII LOCA Analysis was performed using T cold (515 F) for the fluid temperature of the upper hemispherical head. Recent studies and investigations sh'ow that the actual temperature may approach T hot (560 F).

The error was disclosed to be factual by measurement of the coolant temperature in question at another facility.

j 24

AU DI:SIGNATION OF APPAlci:NT CAUSl; OF OCCURRI:MCE Ref:

Intter of August 17, 1976 from Westinghouse to Mr.

V. Stello, Jr.,

(USNPC) "Increaned Temperature of upper head fluid: W submitted to UFC".

AMALYSIS OF OCCURRENCl; 1he error in the assumed upper head temperature primarily affects the Loss of Coolant Accident Analysis (LOCA). The evaluated temperature of 560 F would decrease the cooling effect of the water as it passed through the core in the event of a LOCA.

The reanalysis of the rost limiting LOCA break size assuming T hot (560 F) for the coolant temperature in the upper head disclosed the following:

The peak clad terperature (PCT) for Exxon fuel at 10.15 kw/ft and 180 Effective-Full-Power-Days (EFFD) burnup was calculated to be 1987.5 F.

This number compares conservatively to the previously calculated PCT value of 2019. F for the Exxon fuel at this burnup at 10.5 kw/ft.

In summary, the above results confirm and justify a.85 kw/ft reduction in the limiting LHGR for Exxon fuel for Yankee Rowe.

This penalty of 0.85 kw/ft has been imposed to conservatively counteract the error.

Based upon the above discussion, the potential consequence from the stand-O point of public health and safety will be minimal in that the LHGR has been reduced by 0.85 kw/ft.

CORRECTIVE ACTION A LIIGR penalty of 0.85 kw/ft was imposed to conservatively counteract the error.

This event will be further evaluated to determine the conservatism of the reduction of.85 kw/f t and will be submitted as a change to the Technical Specifications.

FAILURE DATA There have been no previous errors of this nature in the accident analysis.

1. 76-10 DESCRIPTION OF OCCURRENCE On August 30, 1976, an error was uncovered in Figure B-1 of Technical Specifications, Section D.2.c.1 (Figure 3.2-1 of Section 3.2.1 of the Standard Technical Specifications) which delineates the allowable peak Linear lleat Generation Rate (LHGR).

Plant load was immediately reduced

~

by 2.7% to correct the error.

Id 25

DESIGt!ATIO!! OF APPARENT CAUSE OF OCCURRENCE The error in the allowable peak LI!GR was caused by the inadvertent omission of a correction factor to account for the fact that gamma energy is not included in LIIGR limits.

ANALYSIS OF OCCURRENCE The allowable peak LHGR curve, as written, is given in nuclear power and as such does not include a correction for the fact that only 0.973 of the heat was generated in the fuel (proposed Change 125, Table 6-1).

Therefore, when the 0.973 multiplier is applied so that the curve will be consistent with measured numbers, the curve vill be reduced by 2.7%.

Use of the curve without the corrective results in the allowable full power being 2.7% high. The plant was operated with the curve uncorrected for the entirety of Core XII, up to and including August 30, 1976. However, due to the fact that the plant was limited by factors other than LHGR limits, only during twelve of the 221 power operation days was the plant operated in excess of the corrected curve. During these twelve days (which occurred in June, 1976), the actual peak LHGR was between 10.24 and 10.41 kw/ft compared to the corrected limit of 10.22 kw/ft from INCORE run YR-XII-13.

Covering the period of interest, it was found that only 4 fuel rods out of 17,734 in the core were in excess of the corrected LHGR for each of the twelve days.

Therefore, if a LOCA had occurred during those twelve days, the limits set by 10 CFR 50 Appendix K would have been exceeded.

7_U CORRECTIVE ACTION Plant load was immediately reduced by 2.7% in order to correct for the mistake.

A proposed change to the Technical Specifications is being prepared to correct the allowable peak LIIGR figure.

FAILURE DATA This is the first time that this error in the Technical Specifications has 4

been noted.

1. 76-11 DESCRIPTIO!! OF OCCURRENCE l

At approximately 1415, the auxiliary operator noted slight leakage from the weld on the nipple to the No. 3 charging pump discharge header drain valve.

The No. 3 charging pump was isolated at 1500 and repair initiated.

l l

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE The weld failed in the joint between a 1-1/2" x 3/4" 6000# reducing insert and a section of 3/4" Schedule 80 stainless pipe.

It appeared that a

! (~S full 1/16" pullback gap was not allowed during the installation of the

A,/

pipe. The resultant stress rise across the weld caused the weld to fail s

upon experiencing the induced vibrations of the fluid during pump operation.

26 r

1

Oh A!!ALYSIS OF GCCURRE!JCU The failure of the welds did not impose an unreviewed safety question or hazard to the health and safety of the 12ublic in that the leakage was mininal and contained in t he cubicle Area of t.he primary Auxiliary liuilding which is exhausted to the Primary Vent Stack.

Safe operation of the plant was not affected in that two (2) other charging pumps were available for service.

CORRECTIVE ACTIO!J i

The affected section of pipe was cut out and replaced. The plant ensured that the 1/16" minimum pull back gap was maintained.

The welding procedures were reviewed with plant welders to ensure this incident would not be repeated. Since the No. 3 Charging Pump discharge header contains many new socket welds, due to the installation of a pulsation dampener, the system was radiographed, proper pullback was verified on all welds.

FAILURE DATA This piping was a new section installed in the discharge line on May 10, 1976, in accordance with PDC 75-21.

There was one failure of the same type on May 20, 1976.

(See RO 76-05).

1. 76-12 DESCRIPTION OF OCCURRENCE During normal operation the Chemistry Department sampled the Safety Injection Accumulator for boron concentration.

The reralt was 30 ppm over the Technical Specification limit and 80 ppm over the plant's administrative limit.

Higher supervision was notified. The subsequent investigation revealed that this was not a Reportable Occurrence as previously reported.

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE Investigation of the cause by Chemistry Department Supervision revealed r

that on or about 11/16/76, all boron analyses went up %4%.

Further investigation revealed that the standard sodium hydroxide titrant had a concentration of 44% lower than predicted.

ANALYSIS OF OCCURRENCE Since the problem was caused by an error in chemical analysis and not in actual boron concentration, an actual reportable occurrence did not exist.

Therefore, a threat to the health and safety of the public was not created.

CORRECTIVE ACTION All regular boron concentrations were resampled with fresh reagent and found to be satisfactory.

In the future, a known 2300 ppm standard will g-)v be titrated with each safety system sampled.

(,

27

O PAILUIE DATA I4one.

N76-13 DESCRIPTIO!1 OF OCCURREtiCE At 0104 hours0.0012 days <br />0.0289 hours <br />1.719577e-4 weeks <br />3.9572e-5 months <br /> on !!ovember 20, 1976 a plant trip occurred while perforudng OP-4225, Turbine Throttle and Control Valve Surveillance Test.

(See PIR 76-17).

Following the trip, the No. 1 2400V bus section which carries the tio. 2 and 3 MCP's was de-energized per procedure. These MCP breakers, normally, trip automatically on loss of power to the bus.

Upon re-energizing the tio. '. 2400 V bus, the tJo. 3 MCP restarted. Attempts were made to secure the MCP from the control board and locally, but to no avail.

Operations notified Maintenance of the ACB problem 00200 hours.

An electrician manually tripped the AC3 at 0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br />.

DESIG!!ATION OF APPARE!JT CAUSE OF OCCURREtiCE The Westinghouse MCP De Ion Air Circuit Breaker, 50 DH 250-E, trip mechanism had apparently jammed, preventing the trip coil from tripping the breaker, which resulted in the trip coil burning out.

Attempts to manually trip the breaker by the Operators were unsuccessful.

()

!!OTE: For more details see PEL memo to Pile dated 12/17/76, File E-5, 1976.

AtlALYSIC OF OCCURREtJCE The MCP interlocks normally prevent starting a MCP with the TC loop isolation valve open in the respective loop.

However, these interlocks were circumvented in that !!o. 3 MCP ACB failed to open, and the restoration of power energized the MCP.

Ilo reactivity transient occurred since the '_ cop temperature was comparable to the operating loop temperatures.

The isolated loop startup accident analysis assumptions are more limiting than those that existed during this event and assumes that the reactor is at power at the onset of the event. Even then, the analysis for the

'solated loop startup shows that there is adequate margin to DNB and fuel s'elting for the worst possible temperature mismatch between the active and isolated loops. Consequently, even with the most adverse assumptions, n) fuel damage would result for this incident.

Based on the above discussion, the consequence or potential consequence from the standpoint of public health and safety are minimal. Under the worst condition, no fuel damage would result from this incident.

(%)~)

28

e 4

CORRECTIVE ACTION The breaker was removed from service and inspected. A new trip coil was installed and the breaker was tested. The breaker failed to trip once out of 30 trips. This was considered sufficient cause to replace it with a spare breaker, that tested satisfactorily. The Air Circuit Breaker was disassembled and refurbished. No single apparent reason for the binding could be identified. The Breaker will be tested by multiple trips to return to service. Note For more details see PEL Memo to File dated i

12/17/76. File E-5, 1976.

FAILURE DATA 1

i This ACB had a similar malfunction August 2, 1973 with 985 operations recorded on its counter.

It was thoroughly inspected at that time and

{

returned to service. She counter now records 1121 operations, so the ACB has performed correctly 136 times since the previous failure.

Nameplate Data:

j Westinghouse Metal Clad Switchgear, De Ion Air Circuit Breaker, 50 DH 250-E Style No. 23Y 3729-B1, Type Mechanism DH-4E, Date of Mfg. 3/59 Serial No. 3, Breaker Unit 13 Code A.

Max. Design Volt 4760, Amperes 1200, Rated Volt 4160, Cycles 60 Closing l

Volt 125 VDC. Tripping Volt 125 VDC.

4

1. 76-14 DESCRIPTION OF OCCURRENCE During the performance of OP-4204, " Monthly Test or Special Operation of I

the Safety Injection Pumps", Number 1 High Pressure Safety Injection l

Pump's shaft housing seal was found to leak slightly. A maintenance

]

request was initiated to determine the cause of the leak.

1 j

DESIGNATION OF APPARENT CAUSE OF OCCURRENCE l

The cause of the leak was determined to be a cut gasket for the stuffing l

box.

ANALYSIS OF OCCUERENCE i

j Since the leak was small a threat or potential threat to the health and j

safety of the public and plant employees as well as potential damage to

{

other plant components did not exist.

CORRECTIVE ACTION A new-gasket was installed in the pump and the pump was resssembled.

FAILURE DATA This is the-first occurrence of this nature experienced by any of the three high pressure safety injection pumps.

4 1

29

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FUEL PERFORMANCE t

i i

The fuel performance, as measured by reactor coolant Iodine 131 level and I-131/I-133 ratio, is indicative of good fuel with op,1y minor impe r fect j,ons.

The 1-131 level ranged between 6.5 x 10 pc/ml to 1.3 x 10 pc/ml. The corresponding Iodine ratios were 1.12 x 10~

and 6.38 x 10 I

1 i

i l

S l

i i

i i

l l

0 30

/~T Q

SUMMARY

OF COtITAI!1MEllT PE!IETFATICil TESTS 1he following Class B penetration containment leak tests were conducted during this report period:

1.

The Vapor container Electrical Penetrations were satisfactorily tested in accordance with OP-4702, " Vapor Containment Type B and C Penetration Tests", Attachment S.

The blister cover, "0" ring test plug and shell test plugs for the penetration to be tested were removed. An air hose was connected to shell test tap and pressurized to 32 psig.

The hose was disconnected and the rate of decay was monitored. The air hose was connected to the "O" ring test tap and i* was pressurized to 32 psig.

The hose was then disconnected and the rate of decay was monitored. The leakage rate satisfied the c-iterion of the nominal leakage rate.

2.

The vapor Container Personnel IIatch Class B test was satisfactorily conducted in accordance with OP-4702, " Vapor Containment Type B and C Penetration Tests", Attachment W.

The inner and outer seals were checked and the doors tightly closed.

A pressure gage, test isolation valve and service air hose was then connected to the vent at the top of the hatchway. The hatch was pressurized to 32 psig and the pressure decay was monitored for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as well as the initial and final barometric pressure, ambient

()

temperatures and ti.E. for temperatures.

The leakage rate satisfied the criterion of the nominal leakage rate.

O 31

( ~.

d DATA TABULATIONS PLANT STATISTICS YEAR TO DATE Number of hours the reactor was critical 7998 117,398 Poactor reserves shutdown hours 0

0 Ilo'tra generator on line 7887 113,312 Unit reserve shutdown hours 0

0 Gross thernal energy generated (MW11) 4,250,085 59,899,173 Gross electrical energy generated (Fmn!)

1,333,076 18,398,628 IIet electrical energy generated (tant) 1,251,255 17,221,418 Peactor service factor 91.1 81.9 Feactor availability factor 91.2 81.2 Unit service factor 89.8 79.4

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Unit availability factoi-94.9 Not Avail Unit capacity fact.)r (MDC & Design) 81.4 72.4 Unit forced outage rate 6.5 1.5 4

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UNIT SHUTDOW';S & FCR[ DFOWER FEDUCTICNS

(

/

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e Type Forced Duration Reason Method of Corrective Action / Concents Mo.

Date Scheduled (Hours)

(1)

Shutdown (2) 76-1 1/13 F

447 A

3 Electrical fault on #2 reactor coolant po stator 76-2 2/2 F

2.2 A

3 Z-126 line pilot wire relay operaticn at Harriman Station opened line at Yankee, causing loss of l

power to one reactor coolant pulap and auto trip on low flow 76-3 3/28 F

86 A

3 Generator exciter shaft failure 76-4 5/6 S

322 A

2 Install repaired exciter shaft & plug steam generator tubes (2) 76-5 10/30 S

30 A

2 Repair leak in #1 feedwater heater

w 76-6 11/28 F

9.9 G

3 During performance of turbine control and throttle valve surveillance test, local operator lost connunicatio.1s with control room.

Generator load swing resulted in auto trip on high moisture separator level.

(1)

Reason:

(2)

Method:

A - Equipment Failure (explain) 1 - Manual B - Maintenance or Test 2 - Manual Scram C - Refueling 3 - Automatic Scram D - Regulatory Restriction 4 - Other(Explain)

E - Operator Training F - Administrative l

G - Operational Error (explain)

H - Other (explain)

NUMBER OF PERSONNEL AND MAN-REM BY WORK AND JOB FUNCTION

~

Number oTPersonnel '>Too mrem)

To tall 4an-Rem

~

Station Utility montract Work-Sta tion Utili ty Contract Work.

Work & Job Function

__ _Innleyges Aployees ers & Others Employees Employees ers & Others R~ actor Operations &

Surveillance Maintenance Personnel 4

4 3

1.36 1.33 0.62 Operating Personnel 27 0

0 8.87 0

0

' Health Physits Personr.al 8

0 1

2.46 0

0.285 Supervisory Personnel 1

0 0

0.434 0

0.06 Engineering Personnel 0

0 0

0.19 0.123 0

Routine Maintenance Maintenance Personiiel 8

0 4.57 3.31 0.03 Operating Personnel 0

0 0.706 0

0 Health Physics Personnel 0

0 2.35 0

0 Supervisory Personnel 0

0 0.046 0

0.04 0

0 0.020 0.02 0

Engineering Personnel u

Jnservice Inspection Maintenance Personnel 0

0 0

0.124 0.007 0

Operating Personnel 1

0 0

0.360 0

0 Health Physics Personnel 0

0 0

0.182 0

0 Supervisory Personnel 0

0 0

0.002 0

0 Engineering Personnel 0

0 0

0.003 0

0 Special Maintenance tenance Personnel 13 18 4

6.75 6.34 1.37 t,wrating Personnel 0

0 0

0.250 0

0 Health Physics Personne1 6

0 6

1.37 0

0.744 Supervisory Pe 7nnel 2

0 0

0.742 0

0 Engineering Personnel 0

0 0

0.091 0.026 0

Waste Processing Maintenance Personnel 9

5 0

2.60 1.39 0.092 Operating Personnel 13 0

0 3.92 0

0 Health Physics Personnel 5

0 0

0.715 0

0 Supervisory Personnel 2

0 0

0.350 0

0.02 Engineering Personnel 2

0 0

0.602 0

0.008

_ Refueling Maintenance Personnel 0

0 0

0 0

0 Operatirg Personnel 0

0 0

0 0

0 Health Physics Personnel 0

0 0

0 0

0 Supervisory Personnel 0

0 0

0 0

0 Engineering Personnel 0

0 0

0 0

0 TOTAL Maintenance Personnel 40 35 7

15.4 12.38 2.11 0

14.11 0

0 Operating Personnel 43 0

Health Physics Personnel 25 0

7 7.08 0

1.03 Supervisoty Personnel 5

0 0

1.57 0

0.12 ineering Personnel 2

0 0

0.906 0.169 0.008 Grand Total 115 35 14 39.97 12.55 3.27 35

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SUDJECT _1101,IDAYS_ _19D i

The following dates will be paid holidays for Yankee NSD personnel in 1977:

February 21 Monday Washington's Birthday April 18 Monday Patriot's *ay May.30 Monday Memorial Day July 4 Monday Independence Day September 5 Monday Labor Day October 10 Monday.

Columbus Day November 11 Friday Veteran 's' Day November 24 Thursday Thanksgiving December 23 Friday Day before Chris tmas

,_s December 25 Sunday Christmas (Holiday to be celebrated on Monday, December 26).

RS/sb c

Adminis t ra tive[ Supervisor R. Scarfo 4

..