Text

Annual Operating Rept,1977
	ML19340B434
Person / Time
Site:	Yankee Rowe
Issue date:	12/31/1977
From:	; YANKEE ATOMIC ELECTRIC CO.
To:	;
References
	NUDOCS 8011100239
	Download: ML19340B434 (67)
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e REGUTORY DDMT H.E COM 1

- O YANKEE ATOMIC ELECTRIC COMPAhT YANKEE ATOMIC ELECTRIC NUCLEAR STATION ANNUAL OPERATING REPORT FOR 1977 Prepared By Yankee Atomic Electric Company Nuclear Services Division Westboro, Massachusetts 01581 O

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1977 Annual Operating Report

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Table of Contents Title Page No.

Introduction 3

Summary of. Operating Experience 4

i' Amendments to Facility License or Technical 12 Specifications Licensee Event' Reports 15

' Other Changes, Tests and Experiments 30

- A.

Engineering Design Changes 30 B.

Plant Design Changes 31 C.

Tests 38 D.

Summary of Containment Penetration 40 Tests t

Corrective Maintenance Summary -

43 A.

Maintenance Department 43 B.

I&C Department-54 Plant Operating Statistics 59 Histogram 60 Unit Shutdowns & Power Reductions 62 1

Personnel Radiation Exposure 64 Primary Coolant Chemistry 66 E

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INTRODUCTION O

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The Yankee Atomic Electric Nuclear Power Station is a Pressurized Water Reactor (PWR) of 175 MW electrical, maximum dependable capacity (MDC).

The nuclear steam supply system'is a Westinghouse 4 loop reactor. The architect / engineer and constructor for this project was Stone & Webster Engineering Corporation, Boston. - The condenser cooling is once through using the Deerfield River as a cooling madium. The plant is operated in accordance with license DPR-3, ' issued July 19, 1960, pursuant to Docket 4

Number 50-59.

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

This annual report represents the last of a series dating back to 1960. Future reporting to the NRC will be on a monthly basis in a format suggested, in a letter dated September 19, 1977 from the Nuclear Regulatory Commission. This information will be published and be available from the NRC in NUREG 0020, Operating Units Status Report.

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Summary of Operating Experience J

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1-1 At the beginning of the year, the plant was limited to 91.6% power by AP-7104, " Core XII Operational Limits", Revison 26, Attachment A.

1-11 In accordance with Revision 27, of AP-7104, reactor plant power was increased from 550 MW to 580 MW (96.8% power).

1-30 A load reduction to about 85 MWe was begun at 0100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for the purpose of conducting turbine throttle and control valve surveillance testing. Testing was complete at 0520 and the load was returned to the maximum allowable level.

2-3 The plant's power level was reduced to 96.4% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 27, Attachment C.

2-10' The plant's power level was reduced to 96.0% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 27, Attachment D.

2-17 The plant's power level was reduced to 95.6% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the p>.- pose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 27, Attachment E.

2-18 The plant's power level limit was increased to 98.55% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months

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by the procedure AP-7104, " Core XII Operational Limits", Revision 28, Attachment A.

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2-22 The plant's power level was reduced to 98.15% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 28, Attachment B.

2-27 A load reduction to 135 MWe was begun at 0100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for the purpose of conducting throttle valve exercise. The valve exercise was completed and power escalation commenced at 0300 hours0.00347 days 0.0833 hours 4.960317e-4 weeks 1.1415e-4 months . The plant reached its maximum allowable power at 0400 hours0.00463 days 0.111 hours 6.613757e-4 weeks 1.522e-4 months .

3-1 The plant's power level vaa reduced to 97.77% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 28, Attachment C.

3-7 A load reduction to 135 MWe was begun at 0010 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months for the purpose of conducting throttle valve exercise and turbine condenser leak check. The valve exercise leak check were completed and power escalation commenced at 0530 hours0.00613 days 0.147 hours 8.763227e-4 weeks 2.01665e-4 months . The plant reached its maximum allowable power at 0815 hours0.00943 days 0.226 hours 0.00135 weeks 3.101075e-4 months .

3-8 The plant's power level was reduced to 97.38% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-i 7104, " Core XII Operational Limits", Revision 28, Attachment D.

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t 3-15 The plant's power level was reduced to 96.98% at '1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the l

purpose of complying with the operating parameters' of procedure AP-7104, " Core XII Operational-Limits", Revision 28, Attachment E.

3-18 The plant's power level was reduced to 96.67% at 1530 hours0.0177 days 0.425 hours 0.00253 weeks 5.82165e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operating Limits", Revision 29, Attachment A.

3-22 ' The plant's power level was reduced to 96.15% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operating Limits", Revision 29, Attachment B.

3-24 The plant tripped from 181.4 MWe at 0949 hours0.011 days 0.264 hours 0.00157 weeks 3.610945e-4 months , because of a spike in pressurizer level instrumentation resulting from a personnel error during the~ performance of a scheduled surveillance of the Vital Bus inverter (OP-4511). At 1455 hours0.0168 days 0.404 hours 0.00241 weeks 5.536275e-4 months plant recover was held at 44 MWe to allow steam generator chloride concentration to go below.5 ppm.

Plant recovery continued at 2118 hours0.0245 days 0.588 hours 0.0035 weeks 8.05899e-4 months .

3-25 A load reduction for plant shutdown began at 2130 hours0.0247 days 0.592 hours 0.00352 weeks 8.10465e-4 months for the 4

purpose of repairing tube leakage in No. 1 Feedwater Heater. At 2330 hours0.027 days 0.647 hours 0.00385 weeks 8.86565e-4 months the turbine generator was taken off the line.

3-26 At 0310 hours0.00359 days 0.0861 hours 5.125661e-4 weeks 1.17955e-4 months the reactor was critical.

3-27 The turbine generator was phased on the line at 1101 hours0.0127 days 0.306 hours 0.00182 weeks 4.189305e-4 months . A hold occurred at 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months , 48 MWe, to allow steam generator chloride concentration to go below.5 ppm.

Plant recovery continued at 1428 hours0.0165 days 0.397 hours 0.00236 weeks 5.43354e-4 months .

3-29 The plant's power level reached maximum allowable per AP-7104, " Core XII Operational Limits", Revision 29, Attachment B at 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months .

4-2 The plant's power level was reduced to 95.65% at 0000 hours0 days 0 hours 0 weeks 0 months for tha purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 30, Attachment A.

4-9 The pl.?'s power. level was reduced to 94.5% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 30, Attachment B.

4-16 The plant's power level limit was increased to 95.48% at 0900 hours0.0104 days 0.25 hours 0.00149 weeks 3.4245e-4 months by procedure AP-7104, " Core XII Operational Limits", Revision 31, Attachment A.

4-20 Because of indication of excessive air leakage from the Vapor Container, load reduction for a plant shutdown began at 0925. At 1048 the generator was removed from the grid and a plant'cooldown commenced at'2050.

4-22 After a thorough leak detection and.remediation program on Vapor Container continament boundaries a main coolant heat-up was started 5

at-1900 hours.

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'4-23 The reactor was brought critical at 1723, and. physics testing-commenced.

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4-24 Physics testing was completed' at 0425, turbine rolled' at 0855, and generator phased to the grid at 0950.

I 4-25 Load was increased through the day reaching the maximum allowable power level of 94.33%, per AP-7104, " Core XII Operational Limits",

Revision 32, Attachment B.

4-30 The plant's power level was reduced to 93.18% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for_the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits",. Revision 32, Attachment C.

5-7 The plant's power level'was reduced to 92.03% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 32, Attachment D.

5-14 The plant's power level was reduced to 90.87% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 32, Attachment E.

5-19 The plant's power level was reduced to 90.4% at 1600-hours for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 33, Attachment B.

5-21 A load reduction to 90 MWe was begun at 0100 for the _ purpose of conducting throttle valve exercise and turbine condenser leak check.

The valve exercise and leak check were satisfactorily completed and power escalation commenced at 044S hours. The plant reached its 4

maximum allowable power at 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months , t.

5-26 The plant's power level was reduced to 89.93% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 33, Attachment C.

6-2 The plant's maximum allowable power level was reduced to 88.12% at 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months for the purpose of complying with the operating parameters of procedure AP-7104, " Core XII Operational Limits", Revision 33, Attachment D.

6-7 The plant's power level was reduced to 50% at 1845 hours0.0214 days 0.513 hours 0.00305 weeks 7.020225e-4 months as a result of an error in the sma11' break scenario of the Loss of Coolant 4

Accident analysis. (LER 77-30/01T) 6-9 At 1930 a load reduction was started for plant shutdown because of the' uncertainties associated with the small break analysis and the

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generator was taken of f the line at 2045. The turbine-overspeed trip test was conducted at 2110 and the pressurizer solenoid relief valve test was completed at 2252.

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6-10 The reactor shutdown for Core XIII reload was completed at 0045.

After the completion of steam generator safety valve testing, plant 7-S cooldown was started at 0830. At 2020 the Shutdown Cooling System

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was placed in service.

6-11 A Vapor Container air charge was begun at 1201 in preparation for Class A testing and at 1909 the air charge was secured. At 2309 the Class A surveillance test was started.

6-13 Release of the Vapor Container air commenced at 2330 and the Class A test was secured at 2345.

6-14 Completed the air release from the Vapor Container at 0545 (release permit No. 77-54).

At 0910 the pressurizer was cooled down.

The.

missile shield was removed and stored, the equipment hatch cover was removed and both personnel hatch doors were opened.

6-15 Drained Nos. 1, 3 and 4 Steam Generators. All control rod coil stack housings were removed and lead control rod drive housing shield tubes were installed. The Vapor Container manipulator crane load cell and associated electronics were installed. Calibration of the V.C.

manipulator crane was conducted.

6-16 The teleflex frames and tubing were removed from the V.C. and tubes were capped at conoseals.

No. 2 main coolant loop was vented and drained.

6-17 Reactor head studs relaxation was started.

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6-20 All reactor head studs were relaxed and removed. The stacking plate, guide studs, most seal ring, dash pot test stand, and control rod inspection station were installed.

6-21 The fuel transfer upender was installed in the shield tank cavity.

Fuel handling equipment inspections and tests were performed. The reactor head was removed and placed on the flat car.

The equipment hatch was installed and V.C.

integrity set.

6-23 Completed the removal of the guide tubes and drive shafts.

6-25 Conducted a test of the ECCS System to determine the flow characteristics of the various pump and flow path configurations.

No. 2 loop was drained and vented.

6-27 The incore instrumentation package was telescoped and removed. The core barrel was removed and placed on the stacking plate. Loop No.

3 was drained and purged.

6-28 At 1320 fuel movement for Core XIII reload was started.

6-29 The fuel transfer upender was removed from the shield tank cavity, to replace a defective hose, then reinstalled.

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.6-30 All control rods and shim rods were rotated 90 degrees.

7-2 Eddy current testing of Nos. 2 and 3 steen generators' tubes was conducted on 6 percent of the tubes, one tube in No. 3 steam generator was found to have excersive degradation.

7-5 The Z-126 line was removed from service, - for inspection and preventative maintenance, and ther returned to service.

7-8 The drag force testing of the control rods was completed. OP-1700,

" Cycle XIII Reactor Refueling and Component Inspection"'was completed.

The reactor permissive circuit modification (PDCR 75-25) was finished.

The upper core support barrel ~and the instrument package was installed in the reactor.

i 7-13 Charging check, valve CH-V-611A was replaced, radiographic examination conducted and prepared for hydrostatic testing.

7-16 A class "C" test of the Low Pressure Surge Tank was completed. Loop

4 and the safety injection header were drained.

7-18 Completed eddy current testing of No. 3 feedwater heater. Purged and partially drained all four steam generator's secondary side for internal inspection.

7-20 Commenced Safety Injection System modifications (EDCR 77-17 and EDCR 76-6).

7-23 Commenced eddy current testing of No. 2 steam generator tubes.

1 7-25 Completed 100% eddy current and random sludge buildup testing of No. 2 steam generator tubes. Commenced eddy current testing of No.

i 3 steam generator tubes.

1 7-26 Conducted a satisfactory hydrostatic test of No. 2 steam generator's.

secondary side.

7-27 Completed eddy current testing 100% of No. 3 steam generator tubes.

Completed installation of the control rod drive shafts.

7-28 Conducted a satisfactory hydrostatic test of No. 3 steam generator's secondary side. Installed the guide tube, and the core hold-down ring.

7-30 The draining of the shield tank cavity was completed. Pressure-jet cleaning of the service water piping was started. An 80 psig hydrostatic test of No. 4 steam generator's secondary side wac conducted, two tubes that had been plugged previously were found to weep slightly.

i 8-2 The reactor head was removed from the flatcar, and installed on the reactor. Pressurizer internals were inspected and found to be in satisfactory condition. Installed the manways on Nos. 2 and 3 steam l

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Removed guide studs, stud hole plugs, pumped out the stud holes, 8-4 k.

cleaned the stud holes and commenced installing the reactor head studs.

8-6 Completed tensioning the reactor head studs. Conducted a satisfactory Class "C" test of the fuel. chute.

4 J-8-10 Installed new incore thermocouple elements. Tensioning of No. 3 steam generator primary manway was completed.

8-11 Completed stud tensioning on all steam generator manways.

8-12 Completed installing strainers in the check valves of the ECCS lines in preparation for flushing.

8-13 Completed the. flush of the ECCS and removed the strainers.

Hydrostatic test of the ECCS revealed leaky valve bonnets,' valve

. packing leaks, and leaky fittings.

l 8-15 Repaired the leaky components in the ECCS and satisfactorily conducted I

the hydrostatic test of the system. Completed setting the ECCS throttle valves. Drained the ECCS high pressure header, removed the clapper and replaced the cover on SI-V-36.

Ran several flow tests on the ECCS with the check valve clapper' removed.

8-17 Completed the functional test of the safety injection actuation n.-

circuits. Completed the installation phase of the low temperature overpressurization modification.

s 8-23 A hydrostatic test was conducted on the ECCS instrumentation lines to the valve operator. A functional test was conducted on the safety injection accumulator level switches, The pressurizer code safety valves were installed.

8-25 Installed the rod drive cables.

8-26 Established an H2 gas blanket on the Low Pressure Surge Tank. Drew a steam bubble in the pressurizer.

8-31 The reactor rod control precritical check was completed and Core XIII Beginning of Life Zero Power Physics Testing was started.

Functional test of the Intermediate Power Range and Power Range channels was started.

9-2 Completed the low power portion of physics testing. Took the reactor critical and ' commenced drawing steam from the steam system.

9-3 The generator was-phased to the grid at 0630 ending the Core XII-l XIII Refueling and Maintenance Shutdown.

9-4 The plant's power level reached 60% at 0115 hours0.00133 days 0.0319 hours 1.901455e-4 weeks 4.37575e-5 months and was held to 9

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e complete the power coef ficient measurements. At 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months the plant's power level was increased.

9-5 The plant's power level reached 70% at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months and was held to

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complete the power coefficient measurements.

9-6 The plant's power level was increased to 75% at 2130 hours0.0247 days 0.592 hours 0.00352 weeks 8.10465e-4 months , which was the maximum allowable power permitted by AP-7104, " Core XIII Operational Limits", Revision 35 Attachment A.

9-7 The plant's power level was increased to 78% at 2100 hours0.0243 days 0.583 hours 0.00347 weeks 7.9905e-4 months , which was the maximum allowable power permitted by AP-7104, " Core XIII Operational Limits", Revision 36, Attachment A.

9-9 The plant's power level was increased to 82.8% at 0200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months , which the maximum allowable power permitted by AP-7104, " Core XIII was Operational Limits", Revision 37 Attachment A.

The plant's power level was increased to 83.3% at 1200 hours0.0139 days 0.333 hours 0.00198 weeks 4.566e-4 months , which was the maximum allowable power permitted by AP-7104, " Core XIII Operational Limits",

Revision 38, Attachment A.

9-12 A load reduction to come off the line for the repair of loop No.

2 (delta D) flow transmitter at 0845 hours0.00978 days 0.235 hours 0.0014 weeks 3.215225e-4 months . The plant was off the line in Mode 2 at 1048 hours0.0121 days 0.291 hours 0.00173 weeks 3.98764e-4 months . The generator was phased on the grid at 1341 hours0.0155 days 0.373 hours 0.00222 weeks 5.102505e-4 months . A hold at 78.8% for Xenon equilibrium was commenced at 1750 hours0.0203 days 0.486 hours 0.00289 weeks 6.65875e-4 months .

9-13 The plant's power level was increased to 83.3% at 1500 hours0.0174 days 0.417 hours 0.00248 weeks 5.7075e-4 months , which

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was the maximum allowable power permitted by AP-7104, " Core XIII Operational Limits" Revision 38, Attachment A.

9-21 The plant's power level was increased to 84.3% at 1800 hours0.0208 days 0.5 hours 0.00298 weeks 6.849e-4 months , which was the maximum allowable power permitted by AP-7104, " Core XIII Operational Limits", Revision 39' Attachment A.

10-9 A load reduction to 136 MWe was begun at 0100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for the purpose of conducting throttle valve exercises. The valve exercise was satisfactorily completed and power escalation commenced at 0255 hours0.00295 days 0.0708 hours 4.21627e-4 weeks 9.70275e-5 months .

The plant reached its maximum allowable power at 0330 hours0.00382 days 0.0917 hours 5.456349e-4 weeks 1.25565e-4 months .

10-12 At 2300 hours0.0266 days 0.639 hours 0.0038 weeks 8.7515e-4 months a load reduction in preparation for plant shutdown to repair the high pressure turbine casing drain was commenced.

i 10-13 At 0045 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months the generator load was 15 MWe and at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months the plant was operating at maximum allowable power.

10-19 The plant's power level was increased to 84.9% at 1750 hours0.0203 days 0.486 hours 0.00289 weeks 6.65875e-4 months , which the maximum allowable power permitted by AP-7104, " Core XIII was Operational Limits", Revision 40, Attachment A.

11-1 A load reduction of 5 MWe was begun at 0843 hours0.00976 days 0.234 hours 0.00139 weeks 3.207615e-4 months to allow isolation of No. 1 feedwater heater. At 0920 hours0.0106 days 0.256 hours 0.00152 weeks 3.5006e-4 months No. 1 feedwater heater isolated to investigate tube leakage problems.

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11-4 At 0830 hours0.00961 days 0.231 hours 0.00137 weeks 3.15815e-4 months No. I feedwater heater was returned to service, af ter

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tube plugging,.and a load increase to maximum allowable power c ommenced. At 1055 hours0.0122 days 0.293 hours 0.00174 weeks 4.014275e-4 months a plant load reduction was begun to come off the line because of a leak on the No. 4 steam generator blowdown line. The plant was separated from the grid at 1258 hours0.0146 days 0.349 hours 0.00208 weeks 4.78669e-4 months .

11-5

~No. 54 main coolant loop cooldown was commenced at 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months and released to maintenance at 1615 hours0.0187 days 0.449 hours 0.00267 weeks 6.145075e-4 months .

11-7 Af ter repairs on No. 4 steam generator blowdown line were completed a satisfactory hydrostatic test was conducted on the blowdown piping at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months .

11-8 The generator was phased to the grid at 0814 hours0.00942 days 0.226 hours 0.00135 weeks 3.09727e-4 months and reached maximum allowable power at 1058 hours0.0122 days 0.294 hours 0.00175 weeks 4.02569e-4 months .

11-22 The plant's power level was increased to 88.3% at 2300 hours0.0266 days 0.639 hours 0.0038 weeks 8.7515e-4 months , which was the maximum allowable power permitted by AP-7104, " Core XIII Operational Limits", Revision 41, Attachment A.

11-29 The plant's power level was increased et a rate of 1/2% per hour beginning at 1345 hours0.0156 days 0.374 hours 0.00222 weeks 5.117725e-4 months to a maximum allcwable power of 600 MWe as permitted in AP-7104, " Core XIII Operational Limits", Revision 42, Attachment A.

4 12-1 A load reduction from 600 MWt to 584 MWt was performed to remove No. 1 Feedwater Heater from service because of tube leakage problems.

O 12-11 A load reduction to 130 MWe was begun at 0100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months for.the purpose of conducting throttle valve exercises. The valve exercise was 2

satisfactorily completed and power escalation coemenced at 0350 hours0.00405 days 0.0972 hours 5.787037e-4 weeks 1.33175e-4 months .

The riant reached full load at 0500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months .

12-17 A load reduction to come off the line for operator training was begun at 1100 hours0.0127 days 0.306 hours 0.00182 weeks 4.1855e-4 months . The plant was separated from the grid at 1308 hours0.0151 days 0.363 hours 0.00216 weeks 4.97694e-4 months .

After completion of operator training a plant startup was conducted and the generator was phased on the grid at 2247. hours.

4 12-18 At 0050 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months , with plant power at 539 MWt, a hold in power escalation-was begun to allow Xenon equilibrium.

12-19 A load increase to 600 MWt was begun at 0050 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months .

12-31 At year's end the plant was operating at 600 MWt.

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N Amendments to Facility License or Technical Specifications 1.

On March 30, 1977 the Nuclear Regulatory Commission issued Amendment

36 to the Yankee Atomic Electric Company Facility Operating License No. DPR-3. Paragraph 2.C(2) was amended to read as follows:

"(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 36, are hereby incorporated in the license.

The' licensee shall operate the facility in accordance with the Technical Specifications."

The amendment changes certain requirements for the Environmental Monitoring Program specified in Table 4.7-4.

The number of sampling points were increased for air, soil, milk and direct radiation sampling.

Water and river sediment sample points were reduced in number. The sample frequency for soil and river sediment were reduced.

2.

On March 31, 1977 the Nuclear Regulatory Commission issued Amendment

37 to the Yankee Atomic Electric Company Facility Operating License No. DPR-3.

Paragraph 2.C(2) was amended to read as follows:

"(2) Technical Specifications p

The Technical Specifications contained in Appendix A as revised

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through Amendment No. 37, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications."

The amendment will provide limiting conditions for operation up to 500 ef fective full power days (EFPD) for Core XII.

This change involves Figures 3.2-1, 3.2-3, and 3.2-4 which define the. limiting peak linear heat generation rate, the multiplier to account for Xenon redistribution, and a multiplier for reduced power. operation. The previous curves were limited to 389 EFPD. Extending these curves allow for the planned power coastdown at the end of Core XII.

3.

On May 3, _1977 the Nuclear Regulatory Commission issued Amendment No.

38 to the Yankee Atomic Electric Company Facility Operating License No. DPR-3.

Paragraph 2.C.(2) was amended to read as follows:

"(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 38, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications."

l The amendment removed the requirement for a containment entry twice per week for a visual inspection of accessible portions of the Main Coolant System. This requirement became obsolete after completion 12 i

of the initial phases of the Inservice Inspection Program.

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4.

On June 3, 1977 the Nuclear Regulatory Commission issued Amendment No. 39 to the Yankee Atomi; Electric Company Facility Operating License No. DPR-3.

Paragraph 2.C.(2) was amended to read as follows:

"(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 39 are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications."

The amendment incorporated the information contained in Section 218 of the Final Hazards Summary Report to Specification 5.6.1 and also allowed the handling of the spent fael inspection stand and certain fuel handling equipment over the sient fuel pit.

5.

On June 16, 1977 the Nuclear Regulatory Commission issued Amendment No. 40 to the Yankee Atomic Electric Company Facility Operating License No. DPR-3.

Paragraph 2.C.(2) was amended to read as follows:

"(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment 40, are hereby incorporated in the license, r^g The licensee shall operate the facility in accordance with the

(,)

Technical Specifications."

The amendment allows the use of radiation dose integrating devices with an alarm feature or supervision of ongoing work by health physics qualified personnel as options for personnel entry into high radiation Health Physics personnel are not required to have a Radiation areas.

Work Permit (RWP) when entering a high radiation area.

6.

On August 18', 1977 the Nuclear Regulatory Commission issued Amendment No. 41 to the Yankee Atomic Electric Company Facility Operating License No. DPR-3.

Paragraph 2.C.(2) was amended to read as follows:

"(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 41, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications."

The amendment incorporated the revised surveillance requirements to the technical specifications which were made necessary when the hydraulic type snubbers were replaced with the more reliable mechanical type snubbers.

7.

On August 19, 1977 the Nuclear Regulatory Commission issued Amendment

[)

No. 42 to the Yankee Atomic Electric Company Facility Operating License 13

_w k

(

No. DPR-3. -Paragraph 2.C.(2) was amended to read as follows:

"(2) Technical Specifications t

. The Technical Specifications contained in Appendix A, as revised through Amendment No. 41, are hereby incorporated in the license.

l The licensee shall operate the facility in act.ordance with the Technical Specifications."

The amendment adopts the use of Appendix III, ASME Section XI 1974 Edition, Summer 76 Addenda for calibration block design, calibration to notch type reflectors, evaluation of indications exceeding 100% of the reference level, and recording of indications greater than 50%.of the reference' level.

8.

On August 25, 1977 the Nuclear Regulatory Commission issued Amendment No. 43 to the Yankee Atomic Electric Company Facility Operating License No. DPR-3, Paragraph 2.C.(2) was amended to read as follows:

"(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 43, are hereby incorporated in the license, j

The licensee shall operate the facility in accordance with the

. Technical Specifications."

()

The amendment incorporates provisions in the Technical Specifications, required for operation with the refueled Cor" XIII, with an active ECCS accumulator subsystem, and with mclif t:d

.ECCS piping, based on an ECCS performance analysis utilizing certain modeling.

9.

On November 29, 1977 the Nuclear Regulatory Commission issued Amendment No. 44 to the Yankee Atomic Electric Company Facility Operating License i

No. DPR-3.

Paragraph 2.C.(2) was amended to read as follows:

i i

"(2) Technical Specifications 4

The Technical Specifications -contained in Appandix A, as revised through' Amendment No. 44, are hereby incorporated in the license.

The-licensee shall operate the facility in accordance with the Technical Specifications."

l The amendment incorporated the revised Technical Specifications related

~

to-the ECCS accumulator _ actuation time delay setting and to burnup-dependent Linear Heat Generation Rates (LHGR's).

h 14 l

t-

., ~.

s s

Licensee Event Reports n

(~)

The following is a brief description of the 60 LER's which have previously

'^

been submitted to the Nuclear Regulatory Commission. All reports in this series carry the prefix number 50-29/77 to indicate the docket number and year of the event.

01 Failure to Maintain Two Fixed Speed Charging Pumps On January 1, during the performance of corrective maintenance on No.

3 charging pump, No. I charging pump was changed from the fixed speed mode to the variable speed mode.

For a period of seven hours, this condition existed violating Technical Specification 3.5.2.c which requires that at least two fixed speed pumps be maintained during plant operation.

02 Excesnive Leakage from No. 1 LPSI Pump Packing Gland On January 7, Maintenance Request No. 77-14 was generated, to repack the inboard packing gland of No. 1 Low Pressure Safety Injection Pump.

At 1100 hours0.0127 days 0.306 hours 0.00182 weeks 4.1855e-4 months , the Low Pressure Safety Injection Pump was valved and tagged out of service for 'Se required maintenance. During the time the pump was out of service, the three safety injection trains required by Technical Specification 3.5.2.a.2 was not met.

At 1340 maintenance was completed and the pump returned to service.

03 Channel Functional Test Failure - Priniary Vent Stack Particulate Monitor (y

)

On January 10, at 0845, during the channel functional test of the pr. mary vent stack monitor, the particular channel failed to meet the acceptance criteria. A maintenance request was initiated and all planned releases of radioactive material to the vent stack were cancelled. By 0915 the next day, the instrument was repaired and a satisfactory channel functional test conducted.

04 Surveillance Failure - Number 2 Charging Pump On January 14, during the performance of OP-4217, " Charging System Operability Test" Attachment A, Number 2 charging pump failed to meet its Technical Specifications Surveillance Requirements (4.1.2.6.b).

The pump's discharge pressure was not equal to or greater than 30 psig above main coolant pressure. Higher supervision was notified and the decision was made to complete the procedure and restore the charging and volume control system to normal oparation.

05 Inoperative Level Transmitter - Vapor Container Drain Tank Level On January 15, 1977, the Vapor Container Drain Tank was dumped to the Gravity Drain Tank. The dump was secured when the level reached ten inches by remote indication. The indicated level continued to drop to zero inches. The drain tank was then filled to approximately 18 inches and then dumped to ten inches. Once again, the indicated level continued to decrease to zero inches. The tank was then filled to n

15

24 inches. When the supply was secured, the indicated. level continued to rise to 36 inches. At 0040 January 16, a Maintenance Request was initiated to repair the level indication. The investigation revealed a deteriorated diaphragm in the level transmitter and a crud buildup in its sensing lines. The transmitter was replaced with a. spare for temporary use.

The defective transmitter was rebuilt, shop calibrated and reinstalled. The indication was checked with satisfactory results.

06 Boric Acid Mix ~ Tank Flow Path Temperature 4150 F On January 31, at 1100 hours0.0127 days 0.306 hours 0.00182 weeks 4.1855e-4 months , OP-4214, " Chemical Shutdown System Operability Check", was performed. Following the surveillance, two of the nine temperature readout points were found to be below the minimum Technical Specifications requiremeat of 150 F (4.1.2.3.a.2).

07 Accident-Emergency High Level Radiation Monitor Test Circuit Failure On February 4, 1977 a scheduled operational check of the Accident Emergency High Level Radiation Monitor (AEHLRM) was conducted by the plant's Operation Department. During this check, the built-in test source failed to cause the high radiation alarm. The control room operator notified the I&C Department Supervisor, and MR NO. 77-22 was issued.

08/03L Failed Surveillance Test-Primary Vent Stack Iodine Monitor On February 4, during surveillance testing of the primary vent stack iodine channel monitor, it was noted that the count rate did not meet the acceptance criteria when the source was inserted in front of the detector. I&C Supervisor was notified and an investigation was initiated. All planned releases and releases from the evaporator to the atmosphere through the primary vent stack were suspended until the stops for the source drive were aligned and the system performed normally.

09 Primary Vent Stack Particulate Monitor Motor Failure At 0600 hours0.00694 days 0.167 hours 9.920635e-4 weeks 2.283e-4 months on February 21, an alarm was received on the PVS Radiation Monitor instrumentatien. The Shift Supervisor dispatched an auxiliary operator ( AO) to investigate the cause of the alarm. The A0 reported-a blown fuse on the Particulate Channel filter paper drive system.

The Shift Supervisor declared the channel inoperative and notified the I&C Supervisor and a Technician. All planned releases of radioactive material to the-vent stack were cancelled. (A release was not in progress at this time.) Maintenance Request No.77-116 was initiated.

A grab sample from the PVS particulate monitor was analyzed on February 22 and February 23, 1977. The samp'.e was in the form of a fixed filter.

4 The results of the analysis showed that the PVS discharge was within allowable limits.

NOTE: The PVS Particulate Channel is only used as a quick indication k/

16

r

.of release rate increases. Releases are documented using laboratory analysis of a fixed filter.

. /-sg V

The. blown fuse was replaced at 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months to place the channel back in service. On February 23, 1977, the filter paper drive motor was replaced._ At 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months February 24, 1977 a channel functional test was performed, results satisfactory. The channel was declared operable.

10 No. 3 Charging Pump Inlet Line to Relief Valve Failure At approximately 0850 on February 24, while on routine inspection, the auxiliary operator noticed a slight leak on a weld on the discharge line of No. 3 charging pump.

At approximately 0900 hours0.0104 days 0.25 hours 0.00149 weeks 3.4245e-4 months the No. 3 charging pump discharge piping was isolated and drained and made available to maintenance for repair.

. Prior to any repair work, LPE was performed on the suspected area to determine the extent of the failed area.

11 Failed Surveillance - No. 1 Diesel Generator Starter Motor March 1, 1977, 1040 hour0.012 days 0.289 hours 0.00172 weeks 3.9572e-4 months s: During the performance of OP-4207,

" Surveillance of the Station Power System and Emergency Diesel Generators" the number one emergency diesel generator's starter motor failed. The primary side auxiliary operator was monitoring the test locally in the diesel generator cubicle. As the unit was being started, he noticed smoke emanating from the unit's starting motor. The

~ p(_)

auxiliary operator immediately notified the control room operator who-was conducting the test.

The unit was promptly secured from the control room.

12/03L Radiation Monitor Alarm Setpoint Control Failure -#2 SG Blowdown At 0700, 3/6/77, it was noticed by the operating shift that the alarm setpoint adjustment was defective and the alarm could not be adjusted to the correct setpoint. The Shift Supervisor declared the channel inoperative, initiated a Maintenance Request, No.77-154 and notified the I&C Supervisor.

An I&C Technician was notified, who started an investigation at about 0830, 3/6/77. The I&C Technician found that the alarm setpoint control was defective and replaced it with a spare.

The channel.was declared operable at 0900, 3/6/77 and the alarm was functionally tested.

13 Steam Generator C1 Concentration in Excess of Tech Spec Limit On 3/6/77, at 0200 hours0.00231 days 0.0556 hours 3.306878e-4 weeks 7.61e-5 months the plant reduced load to identify and plug a leaking condenser tube. At 0530, the leaking tube had been identified as being in the east box and was plugged. As the plant was increasing load, it was determined that the chloride ion concentration on grab O

17 4

i-e.

sc J

samples of steam generator No. I and 4 blowdown water were 0.51 and

( )'

-0.54 ppm respectively. ~ This is in excess of Technical Specification 3.7.1.6 which allows for a maximum of 0.5 ppa.

By 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months on 3/6/77, the steam generator blowdown had reduced the chloride concentration to 0.15 on S/G #1 and 0.2 on S/G #4.

14/03L Failed Surveillance - Primary _ Vent Stack Iodine Monitor l -

f On March 8, during the performance of OP-4801, " Functional Test of Process Radiation Monitoring System", the _ iodine channel failed to meet the acceptance criteria for source check count rate. A Maintenance Request (77-162) was initiated and I&C Supervision notified. An 4

investigation to determine the cause of the malfunction was started.

The cause of occurrence was due to detector voltage drif t.

Voltage drift is a normally occurring condition caused by changes in the operating characteristics of electronic components.

15 Vapor Container Air Particulate Monitor Failure At 0930, 3/11/77 a reading of the built-in source check was made prior to a V.C. entry. It was. found to be indicating low.

The Shift Supervisor declared the channel inoperative'and notified the I&C i

Supervisor. An I&C Department Technician was dispatched to determine i

and correct the problem.

At_0940, 3/11/77 the I&C Department 1 Technician determined that the

! ()

High Voltage and discriminator control had drif ted out of adjustment.

He immediately' adjusted them to bring the source reading into the acceptable limits.

At 0945, 3/11/77 the Shift Supervisor then declared the channel operable.

The' Main Coolant Leak Detection Vapor Container Air Particulate Monitor failed which violated Technical Specification 3.3.3.1, Table 3.3-4.

16/03L Vapor Container Air Particulate Monitor Failure At approximately 1445 hours0.0167 days 0.401 hours 0.00239 weeks 5.498225e-4 months on March 18, 1977, the vapor container j

air particulate monitor failed. The vapor containment air. particulate monitor is required to be operable per Technical Specification 3.4.5.1.

4

_urab samples were taken and analyzed hourly until the monitor was repaired by 1700 hours0.0197 days 0.472 hours 0.00281 weeks 6.4685e-4 months the same day.

' 17 Steam Cenerator Cl Concentration in Excess of Tech Spec Limit i

Af ter _ the plant trip. at 0945 on March 24, 1977, steam generator blowdown

~

chloride _ concentration increased steadily. Approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the trip chloride ion concentration reached a maximum of 0.75 ppm average. Chlorides. in steam ' generator No. 4 blowdown water were 1.3

. ppm maximum. Subsequently, steam generator blowdown reduced chlorides to less than 0.5 ppm _ by 2045 hours0.0237 days 0.568 hours 0.00338 weeks 7.781225e-4 months the same day. The plant was back 18 i

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at full power by 2245 hours0.026 days 0.624 hours 0.00371 weeks 8.542225e-4 months . Chlorides were reduced to approximately

("')

0.1 ppm by 0800 on March 25, 1977.

v 18 Steam Generator C1 Concentration in Excess of Tech Spec Limit After'the plant maintenance shutdown commencing about 2400 hours0.0278 days 0.667 hours 0.00397 weeks 9.132e-4 months on March 25, 1977, steam generator blowdown chloride concentration increased. Approximately 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months after shutdown, steam generator chlorides were at a maximum with 4 ppm in one steam generator and an average of 2 ppm for the four steam generators.

Steam generator blowdown reduced chlorides to 0.5 ppm by 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months on 3/27/77, about'37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months after shutdown. The plant started up and was at full power by 0200 3/28/77. Chlorides in steam generator water were undetectable by 0800 3/28/77, 19 Loop Seal Radiation Monitor Failure At 0330, April 2, 1977, the control room operator noticed that the pointer on the Loop seal Radiation Monitor was pegged against the alarm setpoint indicator. The panalarm on the M.C.B. and on the Radiation Panel had not operated. The control room operator adjusted the alarm setpoint upscale until it was separated from the meter pointer. The Shift Supervisor initiated MR No.77-223.

At 0515, April 2, 1977, the Shift Supervisor notified the I&C Supervisor and an I&C Technician.

The I&C Technician started to troubleshoot for the cause of the high indication and why the alarm did not operate.

O

\\/

The I&C Technician replaced the Mylar film and protective tape that covers the detector tubes. This action reduced the radiation indication to the background reading.

The I&C Technician cleaned the contacts of the alarm setpoint indication and meter pointer. He checked the operation of the alarm system several times to ensure corrective action and set the alarm setpoint to 1000 cpm.

20 Failure of Air Particulate Monitor Paper Drive Unit At 0830, April 15, 1977 an I&C Technician reported to the I&C Supervisor that the Main Coolant Leakage Air Particulate Monitor (APD) filter tape was not moving. The I&C Supervisor notified the Shift Supervisor on duty who declared the channel inoperative and notified the Health Physics Department and initiated MR No.77-254.

The Health Physics Department initiated hourly grab sampling which was carried out until the channel was declared operable at 1730 hours0.02 days 0.481 hours 0.00286 weeks 6.58265e-4 months on April 15, 1977.

4 The I&C Technician determined that the filter paper drive motor was defective and initiated a' repair.

At 1715 hours0.0198 days 0.476 hours 0.00284 weeks 6.525575e-4 months the filter paper drive system was restored to service 19

r and the Health Physics personnel performed OP-4801, " Fundamental Test

.(-

of Process Radiation Monitoring System" and reported that the channel

- N was operable at 1730 hours0.02 days 0.481 hours 0.00286 weeks 6.58265e-4 months on April 15, 1977.

21 Primary Containment Excessive Leakage The daily air mass calculation performed on April 19, 1977 indicated that a significant leak in the primary containment might exist.

Searches for the source of the leak were begun on April 19 but yielded nothing significant. The April 20 date point of air mass ccafirmed that there was a leak aad of a magnitude such that V.C. integrity had been compromised. In accordance with Technical Specifications (Section 3.6.1.1, 3.6.1.2) the reactor was reduced to less than.00 Kegg at 1050 hours0.0122 days 0.292 hours 0.00174 weeks 3.99525e-4 months , April 20, 1977.

The April 20 data point also indicated that the VC air pressure had fallen below the Technical Specification minimum of 0.75 psi. (Section 3.6.1.7).

' An intensive search for the leak was conducted with no leak found so the reactor was cooled down by 1700 hour0.0197 days 0.472 hours 0.00281 weeks 6.4685e-4 months April 21, also in accordance with Technical Specifications (Section 3.6.1.1).

22 Steam Generator C1 Concentration in Excess of Tech Spec Limit During a plant maintenance shutdown, commencing at about 1030 hours0.0119 days 0.286 hours 0.0017 weeks 3.91915e-4 months on April 20, 1977, steam generator chloride concentration increased.

~ OV Approximately 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months after shutdown, all four steam generators' blowdown chloride concentration reached a maximum of 0.8 ppm.

Thereaf ter, steam generator blowdown reduced dissolved impurities to less than-0.5 ppm chloride by April 23, 1977, when the plant was in I

Mode 5.

23 Waste Gas System Leak On April 25, 1977, a chemist performed a routine waste gas system volume calculation.

It was determined that there was a net decrease in the system volume. Higher supervision was notified.

An auxiliary operator discovered the leaking valve, WD-V-679 at 0700 on April 26.

The valve was repaired by 1100 hours0.0127 days 0.306 hours 0.00182 weeks 4.1855e-4 months that same day.

24 Radioactive Gas Monitor Failure - Loop Seal Discharge At approximately 2230 hours0.0258 days 0.619 hours 0.00369 weeks 8.48515e-4 months on April 28, 1977, the Loop Seal Discharge Radioactive Gaseous Monitor's indication went to zero.

The loop seal discharge monitor is required to be operable when there is radioactive e f fluent in the waste gas surge drum in accordance with Technical

~

Specifications, Section 3.3.3.1 and Table 3.3-4.

The Shift Supervisor notified an I&C Technician and submitted Maintenance Request No.77-284.

25 Radioactive Gas Monitor Failure - Loop Seal Discharge 20

,g e--+

At approximately 0830 hours0.00961 days 0.231 hours 0.00137 weeks 3.15815e-4 months on May 3,1977, the high voltato power supply for the Loop Seal Monitor failed. The loop seal discharge monitor is required to be operable when there is radioactive effluent in the waste gas surge drum according to Technical Specifications 3.3.3.1 and Table 3.3-4.

The Shift Supervisor was notified who in turn, submitted MR No.77-292.

26 Radioactive Gas Monitor Failure - Primary Vent Stack Particulate At 0935, May 18, 1977 the Control Room Annunciator received a Low Flow Alarm on the Primary Vent Stack Particulate Radiation Monitor Channel.

The Shift Supervisor declared the channel inoperative and initiated MR 77-327.

The I&C Department Technician determined that the filter paper was not moving and had plugged the air flow path causing the low flow alarm.

The I&C Department changed the filter paper roll, the filter paper drive motor and the power transistor in the filter paper control circuit. (The last two items were performed as a consequence of a previous malfunction.)

Procedure OP-4801, " Fundamental Test of Process Radiation Monitoring System", was performed and the channel was declared operable at 1450 on May 18, 1977.

27 Low Pressure Safety Injection Accumulator Low Level On May 24, 1977 the I&C Department performed the bi-annual recalibration of the two LPSI accumulator level channels in accordance with OP-6463.

During the performance of the calibration procedure, minor recalibration adjustments were made to bring the channels within procedural limits.

The channels were returned to service. It was noticed-that the actual level in the LPSI accumulator was below the Technical Specification Limit.

28 Radioactive Gas Monitor Failure - Primary Vent Stack Noble Gas At approximately 1935 hours0.0224 days 0.538 hours 0.0032 weeks 7.362675e-4 months , May 30, 1977, the noble gas monitor.of the Primary Vent Stack Radiation Detection System failed low.

This violated Technical. Specification Section 3.3.3.1 and Table 3.3-4.

All planned gaseous releases were secured. The Shift Supervisor declared the channel inoperable.

The Shift Supervisor initiated MR 77-351.

On May 31, 1977, an 1&C Technician trouble shot the system but could find nothing wrong. All voltages were checked and found normal.

At 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months on 5/31/77, OP4801 was performed and the channel was 4

declared operable.

s 21

T) 29 Radioactive Gas Monitor Failure - Primary Vent Stack Particulate b

At 2110, 6/3/77 the Control Room received a Low Flow alarm on the Primary Ven*. Stack Particulate channel. The alarm would not reset and the Shif t Supervisor declared the channel inoperative. Maintenance-Request 77-363 was initiated. The I&C Department Supervisor, Duty Of ficer, and I&C Department Technician were notified.

An I&C Department Technician started to trouble shoot the channel to determine the cause of the malfunction.

The I&C Department Technician found that the flow control setpoints, moveable pointers on the flow indicator / controller, were set too close together. This caused the flow indicator to contact the control setpoints for a time greater than the setpoint of the Time Delay Relay.

The I&C Technician separated and set the flow control setpoints to 70% and 130% of the desired flow.

The channel was declared operable at 2215, 6/3/77.

30/01T Errors in Small Break LOCA Analysis In the course of performing the Loss of Coolant Accident (LOCA) analysis for the Core XII reload submittal, a small break scenario was discovered which appeared mo-e limiting than anay analyzed for Core XII. The plant first reduced power on June 7, to 50 percent but then because of uncertainty with the calculations of header pressure an orderly shutdown was completed on June 10.

31 Radioactive Gas Monitor Failure - Loop Seal Discharge At 1715, June 9, 1977, the Control Room operator noticed that the Loop Seal Radiation Monitor was pegged at the low end of the scale. The Shift Supervisor notified the I&C Supervisor and an I&C Technician.

The Shift Supervisor declared the channel inoperative.

The I&C Technician started to trouble shoot the channel to ascertain the cause of malfunction.

The I&C Technician found that the High Voltage had drifted and was causing the low reading. He adjusted the High Voltage while Health Physics personnel subjected the detectors to a calculated radiation dose.

The Shif t Supervisor declared the channel operable at 1930, June 9, 1977.

32/3L Failure of Containment Isolation Valve to Close At approximately 1300 hours0.015 days 0.361 hours 0.00215 weeks 4.9465e-4 months on 6/11/77, while performing containment leak testing the containment drain line trip isolation valve (TV 209) 22

received a signal to close. Immediate follow-up inspection by plant

(_,s) personnel verified that the valve had not closed, however, within 15 minutes the valve did close by itself. The Shift Supervisor was notified of these events and Maintenance Request 77-389 was initiated.

0 33 Boric Acid Pix Tank Flow Path Temperature.<l50 F At approximately 1800 on 6/20/77 during the performance of OP-4212,

" Chemical Shutdown System Operability Check", it was discovered that 5 of the 8 thermocouple locations were indicating less than 150 F.

0 The lowest temperature was 118 F.

In accordance with Technical Specification 3.1.2.1, the heat traced portion of the flow path must be _> 150 F.

34/03L Pressurizer Code Safety Valves - High Setpoint During the initial set pressure test, PR-SV-181 and PR-SV-182 were found to be out of tolerance. PR-SV-181 was reset to 2481 psig and PR-SV-182 was reset to.536 psig. (Wyle Laboratories Test Report Titled: "Pressurizar Sr 'ety Relief Valve Set Pressure Calibration and Leakage Test Report..

35 Boric Acid Mix Tank-Flow Path Temperature < 150 F At approximately 1400 on 6/27/77 during the performance of OP-4214,

" Chemical Shutdown System Operability Check", it was. discovered that gg 4 of the 8 thermocouple locations were indicating less than 150 F.

()

In accordance with Technical Sptcification 3.1.2.1, the heat traced portion ~ of the flowpath must be > 150 F.

0 36 Boric Acid Mix Tank Flow Path Temperature <150 F At approximately 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months on 8/3/77 during the performance of OP-4214, " Chemical Shutdown System Operability Check", it was discovered that 6 of the 8 thermocouple locations were indicating less than 150 0

F.

The lowest temperature was 78 F.

In accordance with Technical Specification 3.1.2.1, the heat traced portion of the flowpath must be > 150 F.

37 Degradation of Steam Generator Tube (N-26 #3 S.G.)

On July 5, during the performance of scheduled inservice eddy current inspection of Tube N-26 in the No. 3 steam generator, it wac revealed that the tube wall thickness in one location had degraded'approximately 57%, and approximately 40% in another loc. tion.

Both defects were in the OD of the tube and located slightly above the tube sheet. This occurrence is a random defect in a steam generator tube. The tube did not fail in the sense that it lost its integrity. However, the defect is large enough to require that the tube be plugged.

)

38 Boric Acid Mix Tank Flowpath Temperature < 150 F f')

At approximately 1400 hours0.0162 days 0.389 hours 0.00231 weeks 5.327e-4 months on 8/10/77 during the performance of OP-J 23

f3 4214, " Chemical Shutdown Eystem Operability Check", it was discovered V

that 4 of the 8 thermocouple locations were indicating less than 150 F.

The lowest temperature was 111 F.

In accordance with Technical Specification 3.1.2.1, the heat traced portion of the flowpath must a

be >150 F.

39.

Boric Acid Mix Tank Flowpath Not Available At 0030 on July 14, 1977, the gravity feed connection from the BAMT was removed from service to repair CS-V-632, CS-V-633 and CS-MOV-540.

Leakage through these three valves was lowering the heat traced portion of the flow path below 150 F.

See LER's 77-33, 77-35, 77-36 and 77-38.

Also, during this time the charging pump suction header and discharge header division valves were lapped in.

At 2230 on July 16, 1977, charging system and gravity flowpath were returned to service.

Prior to making the system available the following surveillances were satisfactorily completed.

OP-4217, " Charging System Operability Test", Attachments A and C.

OP-4214, " Chemical Shutdown System Operability Check", Attachment B.

40 Low Scram Setpoint - #4 Steam Generator Low Level On July 20, 2030 hours0.0235 days 0.564 hours 0.00336 weeks 7.72415e-4 months , an I&C Technician performing OP-4605, " Low Steam Generator Water Level Channel Calibration No. 3" found that the

, O Low Water Scram Bistable was out of tolerance. The plant was in Refueling, Mode 6, with the shield tank cavity flooded.

41 Failed Surveillance Test - Overheating of #1 Emergency Diesel Generator On August 2, during the performance of OP-4209, " Emergency Diesel Generator Test During Refueling Intervals", Noi 1 Diesel Generator failed to meet the acceptance criteria of Technical Specification 4.8.1.1.2.d.4.

Af ter approximately 25 minutes at > 400 KW load the i

diesel cooling water temperature exceeded the high temperature alarm and continued to increase. The diesel generator was unloaded and secured to prevent damage by overheating.

This occurrence was caused by blockage or flow, due to sludge and scale, in 81 of the approximately 120 radiator tubes. This represents a loss

-in radiator cooling tube capacity of approximately 67%.

42 Failed Surveillance Test - Overheating of #3 Emergency Diesel Generator i

On August 2 during the performance of OP-4209, " Emergency Diesel Generator Test During Refueling Intervals", No. 3 Diesel Generator failed to meet the acceptance criteria of Technical Specification 4.8.1.1.2.d.4.

After approximately 30 minutes at > 400 KW load the diesel cooling water temperature exceeded the high temperature alarm and continued to increase. The diesel generator was unloaded and secured to-prevent damage by overheating.

24 J

l'~)

This occurrence was caused by blockage of flow, due to sludge and scale, in 87 of the approximately 120 radiator tubes.

This represents a loss in radiator cooling tube capacity of approximately 72%.

43 Scram Setpoint Out of Tolerance - Pressurizer Wide Range Level On 9/9/77, during the performance of OP-4626, " Pressurizer Wide Range Level Calibration", it was discovered that the Pressurizer Wide Range Level Scram Setpoint was out of tolerance.

The Technical Specification limit is < 200 inches. The setpoint was discovered to be 206.4 inches. Since the safety analysis (PC-145, Sup. 1 Table 7-2) assumes a setpoint of 209 inches, this occurrence does not present a hazard to the health or safety of the public or plant employees.

44/03L Containment Isolation Valve Failure to Seat On 9/16/77, during Class C testing of valves TV-211 and TV-408, it was observed that the valve leakages were greater than the acceptable limits. The apparent cause of the occurrence was seat and dise deterioration on TV-211 and a hole in the diaphragm on TV-408.

TV-408 had its diaphragm replaced while the seat and disc on TV-211 were machined. After the corrective maintenance was performed, each valve was satisfactorily retested.

45 Main Coolant Flow Trip System Out of Tolerance On August 19 during the performance of OP-4607, " Low MC Flow System A and B (MC Pump Current) Channel Calibration", several of the system overcurrent relays were found to exceed the maximum time delay of 200 m secs. The apparent cause of occurrence was diagnosed as binding between the relay operating plunger and the relay frame.

The overcurrent relays were partially disassembled and the plungers were polished to allow them to operate freely. The system was retested and all relays operated well within specification.

46 Containment Pressure Sensor Out of Tolerance On September 7 during the performance of OP-4634, " Safety Injection Actuation High Containment Pressure Sensors (PS 238 and PS 239)

Performance Test", the setpoint of PS 239 was found to be 5.4 psig.

The Technical Specification is less than or equal to 5.0 psig (Table 3.3-3).

The pressure switch was readjusted to within acceptable limits and is to be replaced in kind if the setpoint drif t reoccurs.

47/03L Safety injection T~

Level Less than 25.5 feet On October 13, the plant had just reached the allowable full load of G

157 MWe following a load reduction for maintenance. A main coolant b

system boration was in progress to compensate for Xenon burnup and 25

(]

to stay within the Rod Restriction Curve. The Safety Injection Tank b

was being utilized as the source of L..ated water, consequently, the tank level was reduced to 25.3 feet. This level was in violation of Technical Specifications 3.1.2.11.b and 3.5.4.a.

There was no significant occurrences that took place as a result of the event.

It is unknown whether or not there have been similar events in the past.

43 No. 2 Battery Charger Bearing Failure During normal operation, at approximately 2315 hours0.0268 days 0.643 hours 0.00383 weeks 8.808575e-4 months on October 16, the Shift Supervisor noted an unusual noise coming from No. 2 battery charger. The duty officer and maintenance personnel were notified.

The subsequent investigation revealed a noisy bearing as the source of the noise. At 0125 hours0.00145 days 0.0347 hours 2.066799e-4 weeks 4.75625e-5 months the battery charger was taken out of service and the No. I battery charger was cross tied to No. 2 battery.

Removal of No. 2 charger and closing the cross tie breaker is contrary to the limiting condition for operations stated in 3.8.2.3.

49 PU-MOV-541 Failure to Close During Surveillance Test On October 19 during normal operation, an operator was performing procedure OP-4212, " Operability Test of the V.C. Recirculation System".

In the course of conducting the procedure, the operator discovered that valve PU-MOV-541 failed to close electrically as required by Technical Specification 4.5.2.b.2.

O V

Upon investigation, it was discovered that the overload heater in the motor operated valve contactor had failed. The failed overload heater was replaced and the MOV was inspected and tested by cycling electrically and by hand.

Since the valve was in its normally open position as specified in Technical Specification 4.5.2.b.2, the ECCS subsystem had an operable flow path as specified in Technical Specification 3.5.2.b.

50 Nitrogen Regulator - Low Setpoint During normal operation, the primary auxiliary operator, on his first round after assuming the watch, observed that the output pressure of SI-PR-59 was 462 ;;ig. Section 3.5.1.e of the Technical Specifications requires that the pressure be maintained at 473 + 10 psig.

Since the pressure was only one psig low and two redundant regulators were available and operating within limits the system would have consequently performed its designed function.

51/03L Radioactive Gas Monitor Failure - Primary Vent Stack Iodine Monitor During normal cperation, while performing Attachment "C" of OP-4600,

" Radiation Monitoring Channel Functional Test", the alarm setpoint n

of the Iodine Channel was found to be set a t 7 50 c pm.

This is a

()

violation of Technical Specification 3.3.3.1, Table 3.3-4.

The 26

background reading was 15 cpm, therefore, the alarm was set to operate at 35 cpm above the Technical Specification Limit of < 700 cpm, greater than the background reading.

,7-V 52 PU-MOV-541 Failure On October 24, during surveillance testing PU-MOV-541 failed to reopen electrically after closure. This occurred while performing valve cycling in accordance with procedure OP-4212, " Operability Test of the V.C. Recirculation System", in conjunction with procedure AP-2005,

" Operations Dcpartment Surveillance Schedule". The cycling of PU-MOV-541 is required at least every 31 days by Technical Specification Sec t ion 4. 5.2.b.i. The valve is required to be operable by the limiting condition for operation (3.5.2.b) of the Technical Specifications.

Upon indication that the valve had failed to open, at 1055 hours0.0122 days 0.293 hours 0.00174 weeks 4.014275e-4 months ,

November 2, 1977, Maintenance Request No.77-852 was initiated to investigate and correct the failure. A previous occurrence of a similar nature occurred on October 19, 1977 which was reported and described in LER 77-49.

53 No. 1 Battery Charger Failure During normal operation on October 20, 1977, while an operator was performing a walk-through inspection, the bearings on the No. 1 battery charger were detected to be noisy. The condition of the bearings were monitored for the next eight working days. On November 5, 1977, during a maintenance outage, the No. 1 battery charger was taken out of service at 0853 for replacement of the bearings. The plant was in Mode 3, hot standby, with main coolant pressure at approximately 970 PSI, and 73 q_)

a temperature of 459 F at the time of the repair. A previous event of a similar nature occurred on the No. 2 battery charger, this was reported on LER 77-48.

The No. I battery charger being out of service in Mode 3 constitutes a degradatior af the No. 1 D.C. distribution train as specified in Technical Specification 3.8.2.3.

54 No. 1 Battery - Low Cell Voltage On December 5,1977, during normal operation, two electricians performed the Quarterly Surveillance Test for Battery Bank No. 1.

This test was performed in accordance with Tech Spec 4.8.2.3.2.b on OP-4501,

" Quarterly Check of the Station Batteries".

In the course of completing this procedure, at approximately 1500 hours0.0174 days 0.417 hours 0.00248 weeks 5.7075e-4 months , battery cell No. 18 failed to meet the acceptance criteria of > 2.1 volts as specified in Tech Spec 4.8.2.3.2.b.

The cell voltage was found to be 2.04 volts and the Shift Supervisor and higher management were immediately notified.

Maintenance Request MR 77-935 and Job Order No.77-265 were initiated.

The battery cell was replaced with a spare cell according to OP-5805,

" Battery Cell-Temporary Replacement", and put on a separate battery charger for 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . The battery cell was allowed to stand idle for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during which its voltage and specific gravity were monitored.

Af ter receiving satisfactory indications, the battery cell was placed

(~}

27 v

on a separate charger for 46 hours5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months . At the end of this 46 hour5.324074e-4 days 0.0128 hours 7.60582e-5 weeks 1.7503e-5 months period, the cell voltage was 2.35 volts and the specific gravity was 1.210.

The cell was then re-installed on December 8,1977 at approximately O

1100 hours0.0127 days 0.306 hours 0.00182 weeks 4.1855e-4 months in accordance with OP-5805, " Battery cell-Temporary Replacement" Part II " Reconnecting the Replaced Cell".

Total time that battery cell No. 18 was out of service was 68 hours7.87037e-4 days 0.0189 hours 1.124339e-4 weeks 2.5874e-5 months . At no time was the battery bank without its normal 60 cell makeup and at no time did the voltage to D.C. Train No. Id rop below 120 volts.

55 Dose Equivalent Iodine 131 > 1 Ci/mi On December 17 during an approximately 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months shutdown for operator training, the dose equivalent Iodine-131 concentration of the primary coolant system exceeded 1 Ci/ml. Technical Specification 3.4.7 limits the specific activity of the primary coolant system to < 1.0 Ci/ gram Dose Equivalent I-131.

The reactor was operating at 100% power prior to the shutdown. Normal l

cleanup flow through the primary coolant purification system was about 25 gpm. There was no degassing operations prior to or during the shutdown.

The dose equivalent iodine concentration exceeded 1.0 Ci/ gram for one continuous interval of 23 hours2.662037e-4 days 0.00639 hours 3.80291e-5 weeks 8.7515e-6 months . The maximum concentration measured was 2.4 Ci/ gram D.E. 1-131 approximately 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months after shutdown.

Thereafter, the primary coolant purification system reduced the D.E.

1-131 concentration to less than 1.0 Ci/gran in 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

4 56 Nuclear Instrumentation Failure O

At 1725 hours0.02 days 0.479 hours 0.00285 weeks 6.563625e-4 months on December 19, 1977, during normal plant operation, No. 5 Intermediate Range Power Channel failed and indication went to zero. After approximately 5 minutes normal indication returned but I

then dropped to zero again. Immediately following the failure of the i

channel the scram logic for the Intermediate Range Power channels was set for single channel operatien in accordance with Technical Specification Table 3.3-1, Action Statement 2.b.

The failed channel j

did not adversely affect normal plant operation. This is the first recorded failure of this nature.

57/03L Low Pressure Safety Injection Pump Packing Gland Failure During the performance of OP-4202, " Monthly Test or Special Operation of the Safety Injection Pumps", at approximately 1003 hours0.0116 days 0.279 hours 0.00166 weeks 3.816415e-4 months on December 29, 1977, the No. 3 LPSI pump outboard shaft packing gland started to overheat, therefore, the pump was immediately shut down. The No.

i 3 LPSI pump was restarted at approximately 1008 hours0.0117 days 0.28 hours 0.00167 weeks 3.83544e-4 months after corrective maintanance had been completed and successfully met the requirements of Technical Specification 4.5.2.a.c.

Technical Specification 4.5.2.a.c requires that each Safety Injection ptsap be operated at least fifteen minutes at least every thirty one days.

28

53 Loop Seal Radiation Monitor - Loss of Power x

At approximately 1600 hours0.0185 days 0.444 hours 0.00265 weeks 6.088e-4 months on December 29, 1977, during normal Mode 1 operation, the loop seal radiation monitor was found to be indicating less than 0 counts per minute. The low indication was found while performing a functional test of the radiation monitor channels in accordance with procedure AP-2007, " Maintenance of Operations Departmental Logs".

The Shift Supervisor declared the channel inoperative, initiated Maintenance Request No. 77-1001 and notified the Instrument and Controls Department. The loop seal radiation monitor is required to be operational by Technical Specifications Section 4

3.3.3.1 and Table 3.3-4.

The low indication resulted from a loss of power to the channel due to a tripped circuit breaker. This is the first occurrence of this nature reported on this channel.

59 Radiation Monitor Failure - Primary Vent Stack Particulate Monitor During normal operation on December 30, 1977, while HP/ Chemistry personnel were investigating cause of a high radiation indication and alarm on the particulate monitor associated with the primary vent stack radiation monitoring system, it was discovered that the filter paper d

was not advancing.

The Shift Supervisor declared the channel inoperative, and suspended all planned gaseous releases.

60 Radiation Monitor Failure - Loop Seal Monitor AV At'approximately 1540 hours0.0178 days 0.428 hours 0.00255 weeks 5.8597e-4 months on December 30, 1977, during normal operation in operational Mode 1, the loop seal radiation monitor was found to indicate less than 0 counts per minute. The low indication found while performing a functional test of the radiation monitor was in accordance with AP-2007, " Maintenance of Operations Departmental Logs".

The Shift Supervisor declared the channel inoperative, initiated Maintenance Request No. 77-1007, and notified the Instrument and Controls Department. The loop seal radiation monitor is required to be operational by Technical Specifications Section 3.3.3.1 and Table 3.3-4.

An investigation was conducted on the channel which revealed a failed Geiger-Muller detector tube. This is the first reported failure of this nature associated with the loop seal radiation monitor channel.

Other Changes, Tests, and Experiments f"'T A.

Engineering Design Changes V

The following is a listing of those fully approved Engineering Design Change Requests (EDCRs) that were implemented during the year 1977.

Engineering Design Change No. 74-3 EDCR 74-3, Filtered Exhaust Ventilation, was completed during the report period. The change consisted of installing two charcoal filter assemblies to filter the exhaust ventilation systems before being discharged to the primary vent stack. This change was made to reduce iodine releases and upgrade the gaseous releases to meet Appendix I philosophy.

Engineering Design Change No. 76-4 Engineering Design Change No. 76-4 entitled, " Mechanical Snubbers",

was completed during this report period. The change consisted of replacing the eight hydraulic snubbers on the pressurizer relief valve lines. This change was made to eliminate the high stress produced by transient dynamic loading during valve actuation.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

(")

Engineering Design Change No. 76-18

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Engineering Design Change No. 76-18 entitled, "Undervoltage Alarm on Emergency Buses", was completed during the report period. The change consisted of adding an additional undervoltage relay on each 480V emergency bus.

A time delay relay was

.ded to actuate an annunciator in the control room. This change was made to ensure that adequate bus vo 'tage is available to successfully operate all of its safety related loads.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Engineering Design Change No. 77-2 Engineering Design Change No. 77-2 entitled, " Control Rod Stack Elevation Change", was completed during the report period. The change consisted of installing a 3 inch spacer around the pressure housing below the flange of the control rod indicating coil stack. This change was made to correct the 3 inch error in the rod position indication system found when the system was changed from incandescent lights with light emitting diodes.

This change does not impose any unreviewed safety question in that A,

30 ix_-

the system will operate in the same manner as covered in the existing safety hazards report.

)

Engineering Design Change No. 77-3

(~'J Engineering Design Change No. 77-3 entitled, " Vapor Container Isolation System", was completed during the report pe riod. The change consisted of replacing the V.C. isolation system solenoids and tubing with larger diameter tubing and specially constructed, larger solenoids. Two solenoids were provided for each trip valve. The change was made to increase the reliability of the V.C. isolation system and meet the intent of the IEEE 279 Standard.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

B.

Plant Design Changes The. following is a listing of those fully approved Plant Design Change Requests (PDCRs) that were implemented during the year 1977.

Plant Design Change No. 72-15 Plant Design Change No. 72-15 entitled, " Low Pressure Safety Injection Pump Mechanical Seals", was completed during the report period. The change consisted of installing mechanical seals on No. 1 LPSI pump.

This change was made to eliminate the leakage along the pump shaf t.

{')N This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing s_

safety hazards report.

Plant Design Char.ge No. 75-11 Plant Design Change No. 75-11 entitled, " Primary Auxiliary Systems Instrumentation Sensing Lines Upgrading", was completed during the report period. The change consisted of replacing the 1/2" stainless steel tubing between the root isolation valves and LLT-221, CH-LD-1, PI-206, Tc-207, PC-205, FI-206; and the LPST Pressure Manifold and PC-201, PC-233, PI-220, CH-PD-14, PS-225; with 3/8" stainless steel tubing, valves, fittings and add a test valve to each instrument.

This change was made to increase the reliability and integrity of the system which decreases the corrective maintenance time and the possibility of a degradation of the systems' boundaries. The addition of the test valves decreases the time necessary for equipment calibration.

This change does not impose any unreviewed safety question in that the system will operate in the same manner so covered in the existing safety hazards report.

31 O

v

ya Plant lesign Change No. 75-22 Plant Design Change No. 75-22 entitled, " Permissive Circuit Change, Substitution of Thermal Converters for Existing Pressure Switches",

was completed during the report pe riod. The change consisted of replacing the permissive initiating pressure switches with three bistable relays. The bistable relays were fed from three thermal converters. The Start-up Rate (SUR) permissive circuit indicating light was rewired to provide an accurate display of the status of the SUR permissive circuit. This change was made to make the permissive circuit'more reliable and to allow the operator to evaluate the state of the SUR permissive circuit.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Lesign Change No. 75-25 I

Plant Design Change No. 75-25 entitled, " Replacement and Relocation of V.C. Isolation Actuation Pressure Switches", was completed during the report pe riod. The change consisted of replacing the Mercoid pressure switches with static "0" ring type switches; to relocate the switches for separation; to add test valves; and to add terminal blocks in the wiring. This change was made to enable the surveillance to be conducted on the V.C. isolation actuation pressure switches.

This change does not impose any unreviewed safety question in that the system will operate in the sane manner as covered in the existing safety hazards report.

Plant Design Change No. 76-5 Plant Design Change No. 76-5 entitled, " Replacement of Diesel Cenerator Control Power Indicating Light Assemblies", was completed during the report period. The change consisted of replacing the lamp sockets and assemblies with ones of a lower voltage rating. This change was made to reduce the possibility of a short in the lamp socket when the lamp burns out, thus making the diesel generator control power more

reliable, This change does not impost any unreviewed safety question-in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-6 Plant Design Change No. 76-6 entitled, " Replacement of the Main Coolant Chargirt Flow Transmitter and Indicator", was completed during the report period. The change consisted of replacing the transmitter and indicator with a Rosemount Model 1151 DP Alphaline delta P Transmitter and a Sigma Model 9262 Indicator / Alarm Unit and an associated power supply. This change was mde to upgrade the system to save time and 32

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material in the miintenance of the channel.

l This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-8 i

Plant' Design Change No. 76-8 entitled, " Stems Generator Wide Range and Narrow Range Level Amplifier Capacitor Change", was completed during t

the repert period..The change consisted of replacing the canned electrulytic capacitors in the steam generator wide and narrow range level anplifiers with discrete components. This change was made because i

the canned capacitors were no longer available.

This change does not impose.any unreviewed safety question in that the system will operate in the sane manner as covered in the existing safety analysis report.

Plant Design Change No. 76-9 i

PDCR 76-9 Installation of Thermocouples of the A.B.M.T. Suction Line, was completed during the report period. The change consisted of installing a temperature monitoring system for the suction line from

~ Boric Acid Mix Tank (BANT) to the charging system. This change was made to permit operation of the ' system under compliance of the Technical Spe ci fications.

This change does not impose any unreviewed safety question in that C

the. system will operate in the sane manner as covered in the existing i

safety analysis report.

Plant Design Change No. 76-11 Plant Design Change No. 76-11 entitled, " Condensate and Feedwater Dissolved Oxygen Monitoring System", was completed during the report period. The change consisted of installing a dissolved oxygen monitoring system to monitor the boiler feed and condensate oxygen concentration. This change was mde to allow continuous monitoring of the oxygen concentration of the steam generator feed to assist the plant in evaluating the secondary chemistry.

This change does not impose any unreviewed safety question in that the system will operate in the sane manner as covered in the existing safety hazards report.

Plant Design Change No. 76-14 Plant Design ~ Change No. 76-14 entitled, " Modifications To The ICI System", was completed during the report period. The change consisted of installing a pressure switch on the CO2 surply to the IC Flux Mapping System transfer device frame and an associated alarm light on the IC

~ Flux Mapping System cabinet in the control' room and an isolation valve 2

33

1 i

l and associated tubing on the leak alarm system of the IC Flux Mapping System. This change was made to save time in performing the surveillance of the leak alarm and to give a warning of low CO2 pressure so the system vill not be operated without the necessary moisture inhibiting gas.

This change does not impose any unreviewed safety question in that the system will operate in the sane manner as covered in the existing safety hazards report.

Plant Design Change No. 76-15 Plant Design Change No. 76-15 entitled, " Spent Fuel Pit Cooler Component Cooling Outlet Flow Indicator", was completed during this report period.

Ihe change consisted of installing a flow indicator in the spent fuel pit (SFP) cooler component cooling discharge line. This change was made to allow more accurate surveillance of such things as: changes in hest load, changes in heat exchanger ef ficiency and changes in the SFP cooling pump capacity.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-16 Plant Design Change No. 76-16 entitled, " Installation of VC Penetration Test aps", was completed during this report period. The change consisted of installing penetration leak test taps on the four lower pd 2400 volt VC penetration cartridges (V.C. Blister 12E, Cartridges 5, 6, 7, 8) and the spare. This change was made to allow testing from inside the vapor container.

This change does not impose any unreviewed safety' question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-18 Plant Design Change No. 76-18 entitled, " Upgrading Distribution Cabinet "A" (EMCC #1)", was completed during the report period. The change 4

consisted of upgrading distribution Cabinet "A" on EMCC No. I from 12 circuit panel to an 18 circuit panel and rearranging the loads a

for a more even current distribution. This change was made to evenly -

distribute the load on each leg of the transformer and provide sufficient spares to meet expanding needs.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-19 34 g

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=..

Plant Design Change No. 76-19 entitled, " Reassignment of No. 2 Charging Pump Power Supply", was completed during the report period. The change consisted of transferring power supply feed of charging pump P-15-2 N

from MCC4 - Bus 1 to MCC2 - Bus 1.

This change was made to provide

('/

power to all three chargirg pumps from a different 480 volt station s_

service switchgear bus to eliminate the possibility of the simultaneous loss of two charging pumps due to a loss of one bps.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the exieting safety hazards report.

Plant Design Change No. 76-21 Plant Design Change No. 76-21 entitled, " Separation of Primary Seal Tank Makeup Pump Power Supply", was completed during the report period.

The change consisted of transferring the power supply of one of the two seal water pumps from MCC 4 - Bus 1 to MCC 4 - Bus 2.

This. hange was made to increase reliability as the loss of one station switchgear bus would not cause the loss of both pumps.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-22 Plant Design Change No. 76-22, entitled, " Changing the Power Supply of CS-MOV-529", was completed during the report period. The change

(~N consisted of changing the power sumply of CS-MOV-529 from MCC 4 - Bus

\\/

2 to MCC 4 - Bus 1.

This change was made to increase plant reliability as the loss of one switchgear bus would not result _in the loss of both motor operated valves (CS-MOV-529 and CS-MOV-540) controlling flow in the boric acid line.

2 This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-23 a

Plant Design Change No. 76-23 entitled, " Replacement of the Boric Acid Mix Tank Heat Trace" was completed during the month. the change consisted of replacing the GE silicone 19 AWG Si 53921 heat tracing with two circuits of Chemflex, each having its own power supply. This change was made to minimize the possibility of the Boric Acid solution temperature going below the 1500F Technical Specification minimum requirement.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety analysis report.

lO l

i

Plant Design Change No. 76-24 Plant Design Change No. 76-24 entitled, " Installation of PAB Heating y

Steam Isolation Valve", was completed during the report period. The

(_,)

change consisted of adding a 6" gate valve in the building heating steam supply line to the PAB.

This change was made to allow isolation of the PAB hcating supply without affecting the availability of the Emergency Boiler Feed Pump.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in th3 existing safety hazards report.

Plant Design Change No. 76-29 Plant Design Change No. 76-29, entitled, " Vapor Container Air Lock Door Interlock Test Switch", was completed during the report period.

The change consisted of installing a key-lock switch in series with the inner hatch door interlock switch. This change was made to allow testing of the interloc circuit to verify that the outer door will not open when the inner door is open.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety analysis report.

Plant Design Change No. 76-31 Plant Design Change No. 76-31 entitled, " Control Rod Position Indicating Coil Assemblies", was completed during the report period. The change

(-~)g consisted of replacing 30 control rod position indication coils wound

(_

on a spool with DGV insulated wire with coils wound on a form with ML insulated wire. This change was made to allow the plant to completely replace one coil stock with a more reliable coil.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 76-39 Plant Design Change NO. 76-39 entitled, "Startup and Power Range Auxiliary Panel No. lY Modification", was completed this report pe riod.

This change modified the Type FN Startup and Power Range Auxiliary Panel No. 1Y to accommodate V.H.S. relays with a coil impedance of 100 ohms.

Since replacement relays for K1109 were no longer available, the change was made so that a replacement panel would be available to maintain proper functioning of the Type FN Startup and Power Range Panel.

This change does not impose any unreviewed safety questions in that the system will operate in the same manner as covered in the existing safety hazards report.

36 O

V

Plant Design Change No. 77-3

/~}

Plant Design Change No. 77-3 entitled, " Safety Injection Tank Heater

\\s Drain Trap Replacement", was completed during the report period. The

~

change consisted of replacing a 3/4" drain trip on the SI tank heating steam with a 1-1/2" trap and associated isolation valves and trap bypass valve. This change was made tohandle the increased flow during cold weather periods. The isolation valves and bypass will allow trap maintenance without endangering the system.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 77-4 Tiant Design Change No. 77-4 entitled, " Spent Fuel Pit Cooling Line Modification", was completed during this report period. The change removed approximately 9 feet from the Spent Fuel Pit cooling return line. This change was made to f acilitate the installation of the new spent fuel inspection station.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 77-5

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Plant Design Change No. 77-5 entitled, " Turbine Throttle Pressure

()

Regulator Sensing Line Removal", was completed during this report pe riod. The change removed and ca r;2d the sensing line from between the north throttle valve and AS-V-689 and the throttle valve pressure regulator. This change allows the plant to remove the throttle valve pressure regulator and also reduce the sources of potential leaks.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

Plant Design Change No. 77-7 Plant Design Change No. 77-7 antitled, " Pump Room Service Water Line Modifications", was completed during the report period.

The change consisted of changing the carbon steel service water lines less than or equal to 2" to copper; replace all branch connections less than or equal to 2" with 2" stainless threadolets; add a tie connection between No. I and No. 2 headers for the service air compressor; and add branch connectors to the 6" lines on the generator and transformer oil coolers for cleaning connections. This change was made to elimiante the added flow restrictions produced by the build up of corrosion products and sludge on the pipe walls.

37

(

)

v

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

. f)

(/

Plant Design Change No. 77-10 Plant Design Change No. 77-10 entitled, " Installation of an Upstream Isolation Valve For TV-410", was completed during this report period.

The change consisted of installing an isolation valve upstream of TV-410. This change was made to allow isolation of TV-410 and its associated piping for repairs while the main steam header is pressurized. This increased overall plant reliability by eliminating the need to shut down to make repairs to TV-410.

This change does not impose any unreviewed safety question in that the system will operate in the same manner as covered in the existing safety hazards report.

C.

Tests A special test was conducted in accordance with procedure OP-2000.43 entitled, "Special Test, Verification of ECCS Flow and Pressure".

The test consisted of determining the two lowest flow trains, determining the highest flow loop with the two lowest flow trains, running all three trains pumping to all four loops, running all chree trains pumping to the highest flow loop, and a repeatability of test results to verify the results of one of the previous tests. This test was performed to determine the actual emergency core cooling system flow characteristics for use in the accident analysis.

This test did not impose any unreviewed safety question not previously evaluated in the safety hazards report.

ECCS.HPSI Flow Test Flow tests of the High Pressure Safety Injection (HPSI) system were conducted. All tests were conducted under OP-2000.45 and were as follows:

Test No. 1 The first test was run to loop No. 4, considering it was one of the closest loops to the header, to determine the actual flow rate through the system. It was also determined during this test that flow to the loop under test resulted in a false pressure (lower by approximately 30 psig) indication and, therefore, a static loop was used to adjust pressure during the remainder of the tests.

(Loops 1 or 4 were used to account for head loss to headers No. 2 and 3).

This test demonstrated that with 150 psig suction pressure to No. I and 2 HPSI pumps and 750 psig on the static loop header, No.1 Loop, the highest flow that could be attained was 315 gpm (375 gpm required).

Test No. 2 38 O

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The second test was then run using No. I and 2 HPSI pumps to No. 1 Loop, reading pressure on Loop 3 and 4 to determine the loop header g-pressure required to attain 375 gpm HPSI Header Flow with 175 psig (s_)S suction pressure on the pumps. The 175 psig suction pressure was determined from the accident analysis which assumes 200 psig suction pressure (reference memo SEG 90/77, P. A. Rainey to H. A. Autio).

From this test it was concluded that the required conditions could not be met by the system.

Test No. 3 On 8/16/77 each HPSI pump was run into a common Loop, Loop No. 1, with 190 + 10 psig suction pressure to attempt to determine the worst flow pumpT's ). The throttle valves were adjusted to approximately 200 gpm to put the pumps on their characteristic curves. The results indicated below the predicted curves for the pumps.

Test No. 4 8/17/77 1230 to 1300 From the results of tests 1 through 3, it was decided to determine what flow could be attained into a loop from two HPSI pumps running with 190 + 10 psig suction pressure and 750 psig loop HPSI pressure indicated on the static loop. The results were still below the acceptance criteria of 375 gpm -- the second run was run with an ultrasonic flow indicator on the LPSI pump discharge header in an attempt to attain a more accurate flow indication -- the indication was too erratic.

f3 Test No. 5 8/17/77 1435

(_)

It was then determined that a more conservative suction pressure on the HPSI pumps would be 160 psig. In order to obtain the worst condition flow, the test was run into Loop No. 3.

The resultant flow at 750 psig was 315 gpm.

Test No. 6 8/18/77 1300 to 1430 After some reanalysis, it was determined that the maximum HPSI pump suction pressure should be adjusted to as close as possible to, but not to exceed 160 psig. The first test, 1300, was run to verify flows at 190 psig suction pressure and the remaining tests were run at <

160 psig suction pressure into each loop. Each test was run for li minutes.

Test No. 7 8/18/77 approximately 1700 This test was run with one HPSI pump alone into one loop in an attempt to determine what the seal leakage was within the pump.

The pump was run into loop 4 and flow and pressure readings were taken. The cooling water to the pump seals was then secured and the pump was run for approximately 30 seconds (27 seconds) under the same conditions. The results showed negligible difference between readings, indicating no 39 p

appreciable seal leakage within the pump.

Test No. 8 8/22/77 1115

(

This test was run to determine the worst flow pump combination into the already known worst flow loop, No. 4.

Resulting in the worst flow combination being HPSI pumps 1 and 3 into Loop 4.

Test No. 9 8/22/77 Test No. 9 was run on each RPSI pump individually to determine the flow characteristics of each pump and compare them against the original The curves indicated a slightly steeper slope than predicted.

curves.

Test No. 10 8/22/77 approximately 1530 Following reevaluation of the data previously taken, accident analysis and pump flow characteristics, it was determined that the best flow for accident conditions would be with two HPSI pumps discharging into 1

one loop at 875 + 50 psig with 175 + 5 psig suction pressure to develop

> 200 gpm flow. The worst flow pumps 1 and 3 were run into each loop, the throttle valves were set for 875 + 50 psig and each loop 200 gpm flow. Pump combinations 2 and 3 and I and 2 demonstrated };into Loop No. 2 to demonstrate 2; 200 gpm flow.

were then run At the completion of test No. 10, the final throttle valve settings were verified and the valves were locked in place.

D.

Summary of Containment penetration Tests A(/

The following containment penetrations were tested.

Title Leak Rate (Wt %/24 hrs.

@ PA=31.6)

Vapor Container Service Water Return Isolation

.01069 Valve TV-408 Fuel Chute Lower Lock Valve

.00026 Vapor Container Open Bulb Leak Monitoring System Could not be Trip Valve TV-211 pressurized Vapor Container Open Bulb Leak Monitoring System

.00000 Trip Valve TV-211 valve Stem Leak-off TV-204

.00000 Vapor Container Heating System Trip Valve TV-409

.00219 Low Pressure Surge Tank and Low Pressure Sample

.00253 40 (m_)

Line TV-213 ECCS Recirculation Header Isolation Valves

.00075 PU-MOV-543 and 544 S

Eight Inch Air Purge Bypass HCV-602

.00000 Isolation Trip Valve TV-212

.00000 Isolation valve CS-V-601 on Cavity Fill Line

.00063 Service Air Supply Isolation Valve to V.C.,

.00249 CA-V-688 Vapor Container Service Water Return Isolation

.01069 Valve TV-408 Fuel Chute Lower Lock Valve

.00026 Fuel Chute Pump-Back System Valves

.00893 V.C. Heating System Trip Valve TV-409

.00219 Main Coolant Drain Isolation Trip Valve TV-202

.01185 Component Cooling Returns Header Trip Valve TV-205

.00016 V.C. Drain Line Trip Valve TV-209

.00016 Electrical Penetrations (Blister 12)

.00332

(

R.;

V.C. Personnel Hatch

.0001 Isolation Valve CS-V-601 on Cavity 1.26 Fill Line Service Air Supply Isolation Valve 4.99 to V.C., CA-V-688 Vapor Container Service Water Return Isolation Valve TV-408 21.37 Fuel Chute Lower Lock Valve 0.53 Fuel Chute Pump-Back System Valves 17.85 Main Coolant Sample Line Trip l'alve TV-206 0.03 Neutron Shield Tank Sample Line Trip Valve 0.0 TV-207 Vapor Container Open Bulb Leak Monitoring 0.004 System Trip Valve TV-211 41 g

I-

' Valve Stem Leak-Off TV-204 0.001 Main Coolant Vent Header Trip Valve TV-203 0.007 Vapor Container Heating System Trip' Valve 4.39 TV-409 Main Coolant Drain Isolation Trip Valve 23.7 TV-202 Stessa Generator ' Blowdown Line Trip Valves 0.024 TV-401A, 401B, 401C and 401D Component Cooling Returns Header Trip Valve 0.31 TV-205 Low Pressure Surge Tank and Low Pressure 5.06 Sample Line TV-213 Vapor Container Drain Line Trip Valve TV-209 0.32 ECCS Recirculation Header Isolation Valves PU-MOV-543 and 544 1.50 Electrical Penetrations (Blister 12) 0.035 Vapor Container Air Purge System 30 Inch 0.0 Inlet Line, TV-4-1 Vapor Container Air Purge System 30" 0.0 Outlet Line, BV-4-2 I

Eight Inch Air Purge Bypasa HCV-602 0.0 Vapor Container Personnel Hatch 0.07 i

Isolation Trip Valve TV-212 0.0 I

Low Pressure Vent Header 0.0 t

4 r

4 42 O

~

COIUtILTIVE NIENANCE SifNARY L

!!AINTEMAriCE DEPARTMENT EFFECT ON CORRECTIVE ACTION TAKES s

SYSTEM ODIPONENT CAUSE RESULT SAFE OPERATION

'IO PREVENT REPETITION

~

waste Line from wastc insufficient Insul-Frozen line None Added insulation to valves Disposal holdup and ation and extreme',y activity dilu-cold weather i

tion decay 4

7 tanks amergency

2 Diesel Defective Improper operation.

None None 4

Generator Generator tachometer of tachometer charging Cap on Dis-Insufficient Seal Leaked None None f

charge Line-

3 Charging Pump

-.g Charging

3 Charging Deteriorated pack-Excessive' Leakage None None Pump' ing and ram charging

3 Charging Deteriorated gasket Leaked None None Pump j

r Safety'

1 LPSI Pump Packing Deterior-Excessive leakage.

None None l

Injection ation

[

i charging 13 Charging Deteriorated packin ) Excessive leakage

'None None Pump and ram l

Charging

1 Charging Deteriorated Valve Noisy None None l

Pump Guide and Seat i

Safety

1 HPSI Pump Defective Gasket Leaked None None i

Injection charging

1 Charging Loss of prime in Pump would not run-None Pump oil drive unit

. None l

1 4

m-----

-w

-+-s-- - -

ur y

S-

\\

CORRECTIVE (tedhTENANCE StBMARY O

~

L)

)

v MAINTENANCE DEPARTMENT ON EFFECT ON CORRECTIVE ACTION TAKEN SYSTDI COMPONENT CAUSE RESULT SAFE OPERATION 10 PREVENT REPETITION I: barging 33 Charging Deteriorated Excessive leakage None None Pump packing t:horging

3 Charging Excessive applied Separated Shrader None None Pump accumulato - torque va7ve from bladder targing

3 Charging Deteriorated Leaked None None a

Pump Gasket

'::arging 43 Charging Cracked pipe Leaked None None Pump

,+

1,a r gi ng Cli-V-693 Deteriorated Leakeq by None None seating surfaces

harging

?'3 Charging Deteriorated Excessive leakage None None

Pump, packing and ram t ir.e rgency No. I Diesel Faulty Starter Diesel did not start None None l'ouer Generator Motor charaing No. 3 Charging Worn packing and Excessive leak rate None None Pump plunger feed and TV-410 Loose bonnet bolts Bonnet leak None None

undensa te morgency No. 3 Diesel Deteriorated gasket Insufficient Pressurc None None

.- e r Generator cn Radiator Cap Seal

CORRECTIVE NTFluNCE S M R i

MAINTENANCE DEPARTMENT M\\LFUNCTION EFFECT ON CORRECTIVE AGION TAKEN SYSTD:

COW GNENT CAUSE RESULT SAFE OPERATION

'ID PREVENT REPETITION Ete rgency-EBFP throttle Misadjusted governor Valve did not close None None

. Coiler Feed valve linkage completedly_

~

Energency EBFP water seal Cracked nipple Leaked None None Boiler Feed to governor

.aste WD-V-679 Deteriorated Leaked by None None Disposal diaphragm tharging No. 3 Charging Insufficiently Minute leak None None Pump torqued head bolts Shutdown Shutdown Coolin ; Worn mechanical Excessive seal leak-None None Ceoling Punp seal age Shutdown Shutdown Cooling Deteriorated threac Coupling leak None None Cooling Pump seal Charging No. 3 Charging Worn packing seals Excessive leakage None None Pump and ram on #3 piston Valve Stem SV-223 Insufficient valve Boron Buildup None_

None Leak Off seal No. 2 Steam Safety Valves Insufficient seat-Small leak None None Generator ing contact tecirculation PU-MOV-548 Defective torque Failed to Operate None None 5

switch i

j Charging Cit-MOV-523 Loose motor lead Failed to open None None i

electrically

%in Steam Trap T-1 Deteriorated body Leaked None None I,,iuxi1 ia ry Dump Line trap Deteriorated Leaked None None Steam strainer cover

p) n

(

CORRECTi m,f.NTENANCE S'GMARY d

4 MAINTENANCE DEPARTMENT EFFECT ON CORRECTIVE ACTION TAKEN SY.T1.21 CG@0.\\BT CA"SE RESULT SAFE OPERATION "ID PREENT REPETITICM Pressurizer PR-M0V-512 Loose valve seat Leaked None None-Cnarging CH-MOV-528 Deteriorated seat Leaked by None None and gate Ci arging CH-M0V-527 Worn seats and gate Leaked by None None C;:ce:ical CS-V-632 Worn internals Noisy operation None None shut 4han CS-V-633 the:aical CS-MOV-540 Deteriorated seats Leaked by None None shutdown and gate D.C. Power No. 1 Battery Worn brushes Short brushes Ncne None g

Charger eeed and HC-V-828 Deteriorated seat

Leaked by None None Ccndensate and disc Charging CH-V-607 Worn seat and disc Leaked by None None.

CH-V-619 4

Ewrgency EG-3ACB Insufficient con-

. Burnt finger None None Power tact pressure B Phase 2 c ar.:

No.'s 1,2 & 3 Broken & cracked Needed rewelding None None re:nera tor S/G Secondary tack welds Side Baffle Plates on feed ring Component CC-V-606 Deteriorated seat Leaked by None None

': val ing and disc Li'ST SV-2150 Deteriorated seat Leaked by None None and disc

O CCaRECT11011NxExaxCESuxxAgx 0

L MAINTENANCE DEPARTMENT M\\LFUNCTION EFFECT ON-CORRECTIVE ACTION TAKEN

_ SYSTD:

CG'@ONENT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITIO'i 4 fety Pipe hangers Unknown Bent eye bolts None None injection Vital Bus Vital Buss Worn brushes Short brushes None

'None Inverter l Sa fety SI-V-39 Packing deterioratec; Gland leakage None None injection

{.

lital Bus D.C. Motor Deteriorated light Light would not work None None Contactor start-socket ing light i

' Main Coolant No. 1 M.C.

Worn internals Noise in clapper None None Check Valve area

,m Charging CH-MOV-526 Deteriorated seats

' Valve did not close None None and gate tight electrically Charging CH-V-611A Deteriorated seat Leaked by None None and disc reed &

Emergency Boil-Packing friction Scored rams None None ecodensatt er Feed Pump

'hemical ICS-MOV-540 Moisture in control Shorted out None.

None thutdown transformer i

- Main Steam No. 2 S/G Non-Steam erosion Hole in body None None return Trap By-1-

pass Valve

' a fety '

SI-V-21 Worn clapper Leaked by None None Injection

'laste L4aste gas surge Dirty Hard to read None None Di.per.a t

,lrum strain.iy M i

tJ ass t

b-O h'

CORRECTIVE hCTENANCE StBolm MAINTENANCE DEPARTMENT l

MALFUNCTION EFFECT ON CORRECTIVE ACTION TXG jfD' CCTONEXT CAUSE RESULT SAFE OPERATION TO PREVEr FIPETITICS cocnent Shield Tank Loose flange bolts Leaked None None c,oling Cavity Coolers

' Min Steam Meter house Deteriorated Blew by None None steam trap internals

); ' tin Coolant No. I loop Original Construc-L.P. indications None None drains tion iapor Clist<

Deteriorated seal Excessive air None None

, 'ontainer 12, Pen ations fi ttings leakage i

5 and 7.

' Pressurizer Surge line pipe Unknown 3 misaligned hangers None None hangers 4: bin Coolant prain line pipe Unknown Locse snuts and one None None

{g hange rs bent rud 4

'ain Coolant

'4 W-326 Tight packing Groaning noise when None None operated I

Primary Pump No.'s 1 and 2 Deteriorated Nitrogen leakage None None cal Water makeup pump cartridges cause pumps to loose suction

'.C.

No. 2 battery Brush rigging Blew charger fuses None None di s tribut ion charger slightly off center

'harging CH-SV-214 Deteriorated seat Leaked by None None and disc

a fe ty SI-V-39, 604, Worr; seat and disc Leaked by None None Injection 68 & TV-605 Wor Personnel hatch Worn seat and disc Leaked by None None con ta i ne r equalizing valve

Il CORPIGIVEGI.NTENANCE SDMARY O

V b,

v PAINTENANCE DEPARTMENT

'E E UIU EFFE C ON CORRE GIVE ACTICN T.UEN SY5 E.':

CGSONE.NT CAUSE RESULT SAFE OPERATION K) PREV 3T REFETITICS S tea:n VD-V-607 Foreign object Valve would not None None Gene ra tor close Component No. I componen'; Worn packing Packing leak None None Cooling cooling pump

~

tontrol Rod Stack No. 24 Open coil Reduced magnetic None None Dri ve field

:Oin VD-MOV-506 Deteriorated seal Oil leak None None j

Coolant l Lr.ergency No. 1 & 3 Scale and sludge Diesel overheated None None j iw.ver diesel gener-buildup ator's radia-gp tor s

I iCharging No. 3 charging Grounded MI cable Electrical ground None None pump motor

..ergency fio. 1 diesel Deteriorated switch No alarm signal None None

) Javie r generator's weiter temper-ature alarm

. l e ty N5-V-13 Deteriorated valve Leaked by None None 1:.jec t i on rh.irging &

Cil-MOV-526 &

Deteriorated seal Leaked None None 1;leed 527 and gasket

,nnaiinq SA-MOV-511 Deteriorated seal Leaked None None and gasket

' hof < lown SC-ifn-551 Deteriorated seal Leaked None None Cooling and gasket

.g r. t y VD-V-859 Unknown Leaked by None None

n.im. t i on

CORRECTIVE

.',TENANCE SUOt\\RY f)

MAINTENANCE DEPARTMENT M1LIEUION EFFEC ON CORREGIVE ACTICS TXG SYS EM CC:00T.T CAUSE RESULT SAFE OPER\\ TION TO PR&BT REPETITICN tustrol Rod Rod Group Selsyn Motor Did not stop at None None Drive Odometer deterioration proper mark at all times Pressurizer No. 2 heater Open heater Reduced heating None None capacity

i uri fication No. 2 purifi-Worn seal Leaked None None cation pump ffoarging No. 3 charging Worn packing and Excessive leakage None None t

pump seals kmm)onent No. 2 Component Worn packing Excessive leakage None None

'ooling cooling pump

.,i Auxiliary AS-V-623,624 Deteriorated seat-Leaked by None None

, i t e rn

& 625 ing surfaces j l'ressurizer PRSH RH4 Unknown Bent eye-bolts hone None q

PRSH RH7 Pipe Hangers

' Pressurizer

8 Heater Loose connection on Tripped on overload None None over load HP Vent Pipe Pipe Hangers Misadjusted hangers Pipe contact where None None

' Drain Hdr.

crossed

N'O CO.TECTIEA1.NTENANCE SDNARY O

J Nj us fMINTENANCE DEPARTMENT

.MMCION EFFECT ON CCRRECTIVE ACTION TAKEN T,.. :

CGCJON%T CiUSE RESULT SAFE OPERATION TO PREVENT REPETITION hin Ccolant fio. 2 ftin Intermittent weep Boron Buildup None None Ccolant Pump tharging No. 3 Charging Deteriorated ram Excessive leakage None None Pump and seals i Charging No. 1 Charging Worn seals Excessive leakage None None Pump I

~~ ponent

!o. 2 Component Worn packing Excessive leakage None None

.coling Cooling Pump i Wutdown Shutdown Cooling Broken cooling Damaged shaft and None None

!.aoling Pump water line bearing L,

s. En rging Spare Relief Noraml wear Required refurbish-None None Val ve

,i ng Ou rging fio. 3 (';arging Worn seais Excessive leakage None None Pump

.:,a r gi nq

o. 3 Charging Worn seals Excessive leakage None None Pump

' hin Coolant tio. 2 Main Worn packing Packing leak None None Coolant Loop Ilypass Vent

'Ioive Charging rio. 3 Charging Worn seals Cxcessive leakage N,.e -

None Pump cononnent

'lo. 2 Component Worn shaft sleeve Increase clearances None None tooling Cooling Pump

~-)

CORRECTIVE F ^ i1X\\NCE SD M\\RY t

~ -

x,i MAINTENANCE DEPARTMENT EFFEG ON CORREGI\\T ACTICN TXG EST'E CCi W NEST l

CAUSE RESULT SAFE OPERATION TO PRE /ENT REPE ITICN C. Er..e r<:en-Na. 's 1 &2 Crack in exhaust Leak None None

-y Power Diesels pipe narqing PU-MOV-541 Open Overload Valve would not None None Heater operate electrically ailiary PC-V-605 Deteriora ted Leak None None

. team Sonnet Gasket

.C.

No. 2 Battery Deteriorated Noisy bearing None None

'i s t ri bu ti on Charge r Bearing

.harging No'. 3 Charging Loose oil line Minute oil leaks None None Pump connections ww N rging SV-212 Deteriorated seal-Leak from bottom None None ing surfaces b6nnet bolts harging No. 3 Charging Worn packing Excessive leakage None None Pump i

M te Gas No. 1 Waste Deteriorated dia-Improper operation Nnne None Gas Compressor phragms and valves sfety SI-V-671, 672 Worn packing Packing leak None None In! ction 673, 674

. uri fica tion PU-MOV-541 Unknown Valve did not None None operate electrically To:vponen t No. 2 CC Pump Worn packing Excessive packing None None

'coling leakage

' man Genera-No. 4 S/G blow-Nozzle and pipe Leak None None for Bl<radown down piping deterioration D.

No.1 Battery Worn Bearing Noisy None None 3a ibution Charger 1

Ent and V D-MOV-505 Tripped overload Would not operate None None

!sain relay electrically

O cc:cecrtvc2^txTexxxcnsetaar (D

~

%\\INTENANCE DEPNGEVf M1LFUNCTION EFFE C ON CORREGIVE ACTION TAKE;

_ SY5Tj'i CGTONENT CAUSE RESULT SAFE OPERATION TO PREVENT R'?ETITION D. C. Dis-Cell IS #1 Unknown Transient Lcw None None trihution Station Bat-Voltage tery Chargin,q 73 Charging Worn Seals Excessive Leakage None None l'tenp

~

N. t. c t..

CS-V-647 Worn Packing Packing Leak None None 1:1icetion and 649 Sa fety No. 3 LPSI Tight Packing Overheated Gland None None inj ect ion Ptetp O

i 9

h

()

Q CORRECTIVE MINTENANCE SGNARY I & C DEPARTMENT EFFECT ON

' CORRECTIVE ACTION TAKEN SYSIDI _

C0fSONENT CAUSE RESULT SAFE OPERATION

'IO PREVENT REPETITION nii.it ion Emergency Test Larnp Voltage Incorrect operation None None mnitoring Gamma Guard drift (in Control Room) fety No. 1 LPSI Meter Movement Sticky operation None None

jection ump Ameter deterioration -

a

-ecd and No. 1 Steam Worn Slidewire Sluggish operation None None an'.k ns a te Generator Wide Brush

' ntrel Range Level Channel E

u d ia t.i or.

Vapor Container Door left open, Short circuited None Reviewed the importance of 1.n i tori ng APD filled with snow closing weather proof doors aith H.P. Technicians i'.L 5 pare Dailey Booster Deteriora ted Incorrect operation None None Relay No. 1 Components l'.C S;ure Bailey Booster Deteriorated Incorrect operation None None Relay No. 5 Components I'c S w e LPSI Pump Deteriorated meter Sticks None None Aicn:e ter movement ~

d '

sa fety No.1 SI Flow Insufficintly Minute leak None None Injection Transmitter tightened fitting

a fety No. '3 SI Flow Insufficiently Minute leak None None injection Transmitter tightened fitting

['.Nuc l ea r Thimble No. 1 Deteriorated No high voltage None None

!ns trumenta tion Source Range detector response f

atety No. 1 SI Loop Dirty Internals Erratic Jperation None None

njection flow ~Trans-mi tter

-l

q(

fl (o)

/

CORRECTIVE E._iTENANCE SD M\\RY I & C DEPARTMENT MALFUNCTION EFFECT ON CORRECTIVE ACTION TNGN J.i. I:.!

COTONtNT CAUSE RESU.'lr SAFE' OPERATION TO PRE E T REPETITION

' rima ry Pressurizer Deteriorated motor Sticky operation None None Level Recorder and worm gear Prima ry Pressurizer Low Faulty Relay Inconsistent reading None None Pressure Alarm Main Coolant No. 1 Main Loose connections Incorrect operation None None Coolant Flow Transmi tter

'la;;o r TV-410 Deteriorated gasket Leak None None

ontainer Isol a tion 4

t.<

FI-SI-6 Meter Worn meter movements Improper operation None None

x tion O

aclear Source Range Weak electron tubes Failed to calibrate None None

!nstrumenta-Channel No.

tion i

hdiation Loop Seal and Deteriorated Improper operation None None ani tori ng Air Ejector Components Power Supply la fety C S-V-655 Deteriorated Leaked None None Injection seating surfaces

-;diation Primary Vent Defective Motor Failed to move paper None None

<:nitoring Stack Gaseous Drive Monitor

' bin Ceolant ib. 3 TC NR Calibration Drift Decrease in temper-None None Ten:pera ture a ture ' indica tion

()

~

(9

\\_/

s,/

()

COPdECTIVE R\\I.'JE RNCE S G MRY I & C DEPARTMENT RiLFUNGIGN EFFECT ON CORRE 51VE ACTION TAKEN S'i'~.~c :

CO M \\H T CAUSE RESULT SAFE OPERATION

'ID PR5 B T REPETITION

.C isolation TV-209 Excessive packing Valve failed to None Counseled plant personnel compression close properly (LER 77-32) 2C Isolation TV-211 Deteriorated seal Leaked by None None ring and plug Service Water PS-409, PIT-414, Deteriorated sensing Possible leak None None PI-414 line

.a fety FI-SI-l Dirty bellows unit Inaccurate indication None None

njection ed and No.1 FeedwaterLoose sealing sur-Improper operation None None 4

s densate Flow Trans-face mitter v

s

.cd & Sleed Vari-orifice Deteriorated motor Leaked by None None operator i'.:J i a ti on Recorder Deteriorated elec-Improper operation None None Ibnitor tronic components

.. leir Channel No. 1 Deteriorated elec-Improper operation None None Instrumenta-tronic components

,:an Main and TV-205 Groove in plug Possible incorrect None None Auriliary operation 5 t ?am thclear SUR Channel unknown Inactive None None Instrumenta-No. 2 tien bin and TV-400 Worn gasket Leaked None None

,axiliary

.> lta.1 m

. - =. - - -

~

C0iCEGIVE t NTENANCE S12ORRY s

9 I & C DEPARTMENT l

J2LFUNCTION EFFE C ON CORREGIVE A GION TAKEN PJSiD CG TC.'E.Y-

.CAUSE RESULT S.\\FE OPERATION TO PREVENT REPETITION feed & Bleed Vari orifice Deteriorated Leaked by None None valve seating surfaces

'la po r FI-SI-5 Fittings galled Leak None None 1:ontainer Steam TV-4018 and Deteriorated Air Leak None

-None.

nora tor TV-401A diaphragm 4

'!; Cooling i.!a ter '

TV-408 Deteriora ted Air Leak None None diaphragm

.5 feedwater CV-1100, Deteriorated Packing leak None None

!w CV-1200 and packing

^

CV-1300

' [ Win.

Pressure Deteriorated Improper indication -

None None lCnolant Indicator (MCB ) components j J eessurizer Pressure Deteriorated Linear Improper operation None None nstrwenta-Channel Amplifier

. tion ervice SW-TV-408 Worn seat and disc Leaked by None None la te r

.a faty LI-SI-4 Deteriorated Improper indication None None 3 1ijection l

movemen ts l

Guclear No. 2 Source Loose Cor.aection

.Irratic Operation None None Ins trumenta tion Range Channel

'! team

30. 2 SG Narrow Unknown Low level alarm ligh t None None I

~

enera tor Range Level would not clear

.harging Flow Indication

D'e'terio ra ted -

Improper operation None None Motor

- d fe ty PS-239 Deteriorated

' Improper operation

-None

,None 16 h.c t ion Switch-l

n n

i )

CCRRECTIVE (

).NTENANCE SG MIRY

(- s) s I & C DEPARTENT

.M\\L!' UNCTION EFFECT ON CORRECTIVE ACTION TA'G

,_ Ei 5]'J_._ CGTONEXT CAUSE RESULT SAFE OPERATION TO PREVENT REPETITICN i)fety No. 1 SI Flow Insufficintly Minute leak flone tione tjection Transmitter tightened fitting Sa fet y PS-233 6 Galled Threads Valves difficult None "one inj ee: ion PS-239 Test to operate Valves Charuing

low Transducer Dirty Slidewire G Erratic Operation None None Low Gain Waste l'RC-303 Control Linkage Not maintaining None None

!)i:-posa l Binding correct discharge pressure en Air SOV-236 Norn Latch Spuro,us Trips None None t

~

.- t.y e

,7 i

) -

i l

Plant Operating Statistics l

1 Year

' Cumulative Number of hours reactor critical 6,595 123,993 i

Hours generator on line' 6,471 L19,784~

i Gross Thermal energy (MWH) 3,516,596 63,415,768

. Cross Electrical generation (MWH) 1,091,719

.19,490,347 Net electrical generation-(MWR) 1,025,170 18,246,588 Unit se rvice factor 73.9 79.7 1 -

Unit availability' factor 73.9 79.7 Unit capacity factor (MDC& DER) 66.9-72.1 Unit forced outage rate 2.9 1.6 2

4

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UNIT SHUTDOWNS AND POWER REDUCTIONS i-t TYPE METHOD OF F-FORCED DURATION SHUTTING DOWN NO.

DATE S-SCHEDULED (HOURS)

REASON THE REACTOR CORRECTIVE ACTIONS / COMMENTS 770130 S

B 1

Load reduction to 50% for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for tarbine throttle valve surveillance testing.

770227 S

B 1

Load reduction to 75% for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for turbine throttle valve surveillance testing.

i 770307 S

A,B 1

Load reduction to 50% for 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months for turbine throttle valve surveillance testing and condenser inspection.

77-1 770324 F

4.32 G

3 False scram signal, pressurizer low level, resulting from low voltage on vital bus.

e to t

77-2 770325 S

35.52 A

1 Plug 7 leaking and 8 surrounding tubes,

1 FW heater.

i 77-3 770420 F

94.0 A

1 Indication of excessive containment leakage.

770521 S

A,B 1

Load reduction to 50% for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for turbine throttle valve surveillance testing and condenser tube plugging.

770607 S

F 1

Load reduction to 50% for evaluation of ECCS model 77-4 770609 S

2049.75 C

1 Core XII-XIII refueling.

77-5 770912 F

3 A

1 Repair leak on loop flow AP cell.

f 771013 F

A 1

Load reduction to minimum load for 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months to repair leak on h.p. turbine casing

n im m

U C

(d UNIT SHUTDOWNS AND POWER REDUCTIONS (Continued)

TYPE METHOD OF F-FORCED DURATION SHUTTING DOWN MO.'

DATE S-SCHEDULED (HOURS)

REASON THE REACTOR CORRECTIVE ACTIONS /COMFENTS 77-6 771104 F

92 A

1 Repair leak-stear:: generator blowdown line inside containment.

771211 S

B 1

Load reduction to 75% for 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months for turbine throttle valve surveillance testing.

77-7 771217 S

10 E

1 Reactor operator training

p.

p-.

p L-)

YANKEE ROWE PERSONNEL AND MAN-REM BY WORK AND DUTY FUNCTION REPORT

- - 1977 ***

End of 4th Quarter

--Number of Personnel Over 100 MREM-

-Total Man-Rem =

Station Utility Contract Workers Station Utility Contract Workers Employees Employees and Others Employees Employees and Others Reactor Operations & Surveillance 8

2 0

2.306

.489

.297 Maintenance Personnel 8

2 0

2.306

.489

.297 Operating Personnel 24 0

0 9.139

.000

.000 Health Physics Personnel 6

1 10 7.581

.444 2.332 Supervisory Personnel 0

0 0

.195

.090

.318 Engineering Personnel 1

2 0

.749

.459

.038 Routine Maintenance Maintenance Personnel 19 36 11 7.900 13.381 5.425 Operating Personnel 2

0 0

.816

.000

.000 Health Physics Personnel 3

C 1

1.196

.000

.407 Supervisory Personnel 0

1 2

.057

.255

.983 Engineering Personnel 0

0 0

.007

.303

.010 Inservice Inspection os Maintenance Personnel 6

26 16 1.451 13.774 9.128 Operating Personnel 4

0 0

3.763

.000

.000 Health Physics Personnel 1

0 0'

.518

.020

.300 Supervisory Personnel 5

5 2

4.102 1.669

.882 Engineering Personnel 3

3 4

2.985 1.963 2.310 Special Maintenance Maintenance Personnel 20 47 73 16.098 24.688 77.120 Operating Personnel 1

0 0

.627

.000

.000 Health Physics Personnel O

11 x 227

.000 8.656 Supervisory Personnel 2

1 20 1.lta

.357 17.750 Engineering Personnel 1

5 1

.129 2.475

.122 Waste Processing Maintenance Personnel 0

1 0

.237

.315

.109 Operating Personnel 6

0 0

1.565

.000

.000 Health Physics Personnel 1

0 7

.195

.058 4.095 Supervisory Personnel 0

0 1

.000

.805 Engineering Personnel 0

0 0

.000

.002

J

{

YANKEE ROWE PERSONNEL AND MAN-RD1 BY WORK AND DUTY FUNCTION REPORT

- - 1977 ***

End of 4th Quarter

Number of Personnel Over 100 MREM Total Man-Rem Station Utility Contract Workers Station Utility Contract Workers Employees Employees and Others Employees Employees and Others Refueling Pbintenance rersonnel 17 34 12 6.612 14.197 7.695 Operating Personnel 23 0

0 10.134

.000

.000 Health Physics Personnel 5

1 19

.912

.107 14.937 Supervisory Personnel 2

2 0

.716

.825

.420 Engineering Personnel 1

2 0

.264

.237

.043 Totals Maintenance Personnel 70 146 112 34.604 66.844 99.774 Operating Personnel 60 0

0 26.044

.000

.000, Health Physics Personnel 23 2

48 13.629

.629 30.727 Supervisory Personnel 9

9 25 6.253 3.196 21.158 Engineering Personnel 6

12 5

4.134 5.437 2.525 Grand Total 168 169 190 84.664 76.106 154.184

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