Text

Seimiannual Operating Rept,Jul-Dec 1975
	ML20030A487
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	12/31/1975
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
References
	NUDOCS 8101090728
	Download: ML20030A487 (87)
	v • d • e

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CONSUMERS POWER COMPANY Docket 50-155 License DPR 6 23RD SEMIANNUAL REPORT OF OPERATIONS OF BIG ROCK POINT PLANT July 1, 1975 - December 31, 1975 I.

INTRODUCTION - SEMIANNUAL OPERATING REPORT The 13th fuel cycle continued smoothly, except for minor outages, throughout the entire reporting period. As a result, the reactor availability factor was 98.2% and a turbine generator availability of 97 7% (availatilit,ies based on operating units status report definitions). The average grosi generation was 57.6 MWe at an average themal power.'.evel of 1831 M5:t.

The total gross generation for this period was 254,400 MWhe with a gross unit efficiency of 315%.

i II.

OPERATIONS SIN.ARY A.

Changes in Plant Design Facility changes are as follows:

1.

C-272 This Facility Changc consisted of replacing the motor operating switches BMC-5530 and BMC-5531 on MO-7073 and MO-70Th with Remote Manual Control i

switches. The new switches have the " lock out" feature, thereby making these control switches compatible with all present FMC switches installed on the r nels in the Control Room.

2.

C -30)

This change consisted of replacing the 138 kV line voltmeter with a new unit (digital voltmeter). The original indicator had become sluggish and insensitive to minor changes in voltage level. The digital voltmeter will provide the sensitivity required to maintain desired voltage schedules.

3 C-310 This change involved returning the stairvell lighting on the south side of the Service Building to Lighting Panel 2L-16 from 5L 4.

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C-313 This Facility Change consisted of separating the power to the off-gas filter cell heaters, from the isokinetic probe heater, vest side lights and receptacles, east side receptacles and lights, and off-gas sampler heater. Panel 6L Breaker 9 now has the off-gas filter cell heaters. Breaker 1 has the east side receptacles and lights, vest side receptacles and lights, off-gas sample heater, and isokinetic probe. The change was for personnel safety.

5 C-317 This Facility Change provided a temporary power source for the con-struction activities for the RDS System. The power load was added to the secondary side of station power transformer #7 and fed by two temporary 100 kVA transfonners.

6.

C-319 This change consisted of installing an outside water tap on the well water line that runs through Warehouse #1 and Office #1. The tap was necessary in order to supply water to the construction Company working on the RDS project.

7 C-320 This Facility Change consisted of supplying electric power end gas for two office trailers that were moved on site. A pole was set and a trench was dug to the west of the Service Building Aanex. Power came frcxn the 75 kVA transformer next to the Service Building Annex, and the gas line connected to the gas service that supplies the Service Building Annex.

Perfod}jce Characteristics B.

At the stfrt of the report period, the reactor thermal power output was being maintcined at 199 Mwt with a gross electrical generation of 63 Mwe. In addition, the off-gas re' ese rate was ~2,500 pci/sec.

While conducting the daily control rod drive exercise test on July 25, E 4 drive would not return, upon a withdrawal signal to its fully withdrawn position, 23 Attempts to withdraw the rod only led to its 2

insertion to Position 19 At this point E 4's selector valves were valved out. However, the rod's scram capability was not affected.

A review of the core physics parameters showed a possible encroach-ment of the MAPLHGR limits on F-type fuel with the most conservative fluxwire correction. As a result, power was reduced to 198 I 2 Mwt.

A main condenser tube leak on August 21 resulted in an increase in the reactor water conductivity. This tube leak did not present any operational problems until November 13 when the unit was shut cdwa for a scheduled condenser tube plugging outage.

Incore detector 13B reading increased on September 14 coincident with one notch withdrawals of control rods B 4 and E-3 At this point intermittent "incore flux high" alarms esme in for 13B. The detector did not resp *ond when rod E-3 was rel.icerted to its previous position.

Further investigation resulted in taking the detector cat of service.

This left eighteen out of twenty-four detectors available.

A fuel shipment containing seven mixed oxide assemblies from Exxon Nuclear C ;mpany. Inc. arrived on site Tuesday, September 16, 1975 The fuel containers were unloaled and the fuel wc; stored in New Fuel Storage. These mixed oxide fuel assemblies were subjected to a final on-site inspection by the plant staff in conjunction with an Exxon QC inspector. No anomalies were noted.

Another fuel shipment of twenty urania fuel assemblies arrived on site on September 25, 1975, and are now being stored in the containment inside their shipping containers.

On September 25. an investigation into the cause of a 480-V ground indication led, via the elimination process, to the removal of the No. 2 condensate pump from service. This necessitated a power re-duction to 4 50 MWeg (one reaator feed pump). The ground was found l

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insertion to Position 19 At this point E 4's selector valves were i

valved out. However, the rod's scram capability was not affected.

A review of the core physics parameters showed a possible encroach-ment of the MAPLHGR limits on F-type fuel with the most conservative fluxwire correction. As a result, power was reduced to 19812 MWt.

A main condenser tube leak on August 21 resulted in an increase in the reactor water conductivity. This tube leak did not present any operational problems until November 13 when the unit was shut down for a scheuuled condenser tube plugging outage.

Incore detector 13B reading increased on September 14 coincident with one notch wi* hdrawals of control rods B 4 and E-3 At this point intemittent "incore flux high" alams came in for 13B. The detector did not resp'ond when rod E-3 was reinserted to its previous position.

Further investigation resulted in taking the detector out of service.

This left eighteen out of twenty-four detectors available.

A fuel shipment containing seven mixed oxide assemblies from Exxon Nuclear Company Inc. arrived on site Tuesday, September 16, 1975 The fuel containers were unloaded and the fuel was stored in New Fuel Storage. These mixed oxide fuel assemblies were subjected to a final on-site inspection by the plant staff in conjunction with an Exxon 4C inspector. No anomalies were noted.

Another fuel shipment of twenty urania fuel assemblies arrived on site on September 25, 1975, and are now being stored in the containment inside their shipping containers.

On September 25. an investigation into the cause of a 480-V ground indication led, via the elimination process, to the removal of the No. 2 condensate pump from service. This necessitated a power re-duction to +.50 MWeg (one reactor feed pump). The ground was found

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to be in a viring junction box contact on the No. 2 condensate pump motor. Reactor power output was increased to 197 MWt following corrective maintenance on September 27 The semiannual containment component leak rate test (exclusive of fluid containing isolation valves) was performed on September 22 through September 27 Data analysis indicated a total leak rate of a-124 lbs/ day or about 46% of the 10 CFR 50 Appendix J acceptance criterion which was less than 27% of the Technical Specification limit.

P The No. 2 recirculation pump inner seal showed a sharp rise in tem-perature and a decrease in shaft seal pressure drop several times during this reporting period. These fluctuations occurred on July 2, September 2l October 11, November 5 and December 25 The reactor operators were able to decreas, the pump shaft.eal temperature and stabilize the shaft seal. Pressure dropped within the corrective action limitations.

The plant load was reduced to 30 Mweg on October 19 for modifications l

to the Livingston Substation. While load was at 30 MWeg, the turbine bypass valve opened partially and load dropped to 8 MWeg. The Initial Pressure Regulator (IPR) control had failed at this point. The turbine governor control was also found to be unresponsive and a check of the turbine controls revealed steam to be blowing out of the IFR mechanism.

The IPR was tripped, the steam leak isolated and load increased to 30 MWeg where it was maintained until October 20 when load was increased to 52 MWeg. The load was carried at 52 MWeg with synch-governor control until October 24 when IPR control repairs were completed and load was increased to 61 MWeg (~ 192 MWt) which was the current thermal power limit based upon the hAPI11GR Technical Specification.

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At 0140 hours0.00162 days 0.0389 hours 2.314815e-4 weeks 5.327e-5 months on October 30 the IPR failei, again dumping the load from 29 MWeg to 5 Mbeg. Turbine regulation was performed with only the synchronizing governor and the IPR cont ol system was tripped.

Load was then returned to 26 MWeg and a power reduction rate of 1 MWe/6 min. was initiated until 5 MWeg was obtained. At a power level of 5 MWeg, the turbine generator was taken off the line and vl75,000 lbm/hr of steam was bypassen to the condenser. The turbine generator was then tagged out of service at 0407 hours0.00471 days 0.113 hours 6.729497e-4 weeks 1.548635e-4 months to repair a steam leak on the stage drain line from the high-pressure turbine to the high-pressure heater.

t l

1 The turbine generator was returned to service at 0948 on October 30-and the load was inc; eased to a maximum of 52 MWeg because the in-stallation of a valve bellows in the IPR control system was not I

completed. 'The turbine was placed on IPR control at 2321 hours0.0269 days 0.645 hours 0.00384 weeks 8.831405e-4 months on October 30 and plant load was increased to 60 MWeg.

l The unit was on-line until 0452 hours0.00523 days 0.126 hours 7.473545e-4 weeks 1.71986e-4 months on November 13 It was then taken off-line and the reactor taken suberitical, for a scheduled 48-hour outage, to plug leaking tubes in the main condenser. The leaks in the condenser had degraded to the point that the condensate demineralizer resins were being exhausted at a higher than normal rate. Three tubes were found to be leaking on the north tube bundle and vere subsequently plugged. The turbine generator was returned to service and was on tne line at 0155 hours0.00179 days 0.0431 hours 2.562831e-4 weeks 5.89775e-5 months on November 15 During the condenser tube repair outage the reactor was maintained suberitical with 4-251 notches withdrawn. On November 13 at 1321 hours0.0153 days 0.367 hours 0.00218 weeks 5.026405e-4 months , during a routine instrument check of Channel 1 picoammeter with Channels 2 and 3 P coammeters above the downscale trip, the i

reactor protection Channels 1 and 2 tripped causing a control rod scram. Indication on control rod E k was lost at this point.

(The rod had been considered stuck as to withdrawal at Position 19.)

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At 1550 hours0.0179 days 0.431 hours 0.00256 weeks 5.89775e-4 months indication was regained and verification was made that the rod had scrammed. Position indication to E 4 control rod drive was again lost on November 16 and attempts to regain indication were unsuccessful. The drive was then valved out (withdrawal signal only) at the fully inserted position and a manual scram of the drive was perfomed daily.

As power was being increased on November 19, at 171 MWt, the reactor feedvater flow began oscillating causing the picoammeters to swing over the 100% setting. The subsequent recalibration of the p.'coam-meters set the 100% power reading at 175 MWt due to MAPLHGR limitations on F-type fuel. The power was then increased to 172 MWt.

All control rods were fully withdrawn, except E 4 being fully inserted,

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on December 1 at 1541 hours0.0178 days 0.428 hours 0.00255 weeks 5.863505e-4 months .

The weekly pipe tunnel inspection of December 2 discovered a steam leak coming from under the front end of the turbine. Due to the amount of steam leaksge, a lead reduction began at 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months on December 3 Leakage was due to a hole at a velded, junction of the high-pressure turbine casing reducer and a 1/2-inch drain pipe from the No. 2 stage drain header. Attempts to install a temporary patch on the pipe were not successful and load was increased to 53 MWeg.

The unit was taken off the line at 0333 hours0.00385 days 0.0925 hours 5.505952e-4 weeks 1.267065e-4 months on December 6 and the reactor was suberitical at 0342 hours0.00396 days 0.095 hours 5.654762e-4 weeks 1.30131e-4 months , in order i,o perfom the semi-annur.1 control rod drive testing and to repair a steam leak in the high-pressure turbine casing reducer. As described above, the unit was off the liu for 49 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and was returned to service at 0516 hours0.00597 days 0.143 hours 8.531746e-4 weeks 1.96338e-4 months on December 8.

All rods were fully out, except E-4 which was fully inserted, for the remainder of the reporting period. The gross electrical output at the end of the reporting period was 50.4 MWe at a themal power i

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.m.

of 158.1 MWt. The off-gas release rate was 3,600uCi/seewhichwas about1,100uCi/secgreaterthantheoff-gasreleaserateatthe l

beginning of the reporting period.

C.

Changes in Procedures Manual Which Were Necessitated by Preceding A and B or Which Otherwise Were Required To Improve the Sa?cty of Facility Operations The following procedural changec were made wit.h respect to plant operations:

Change No.

Section and Systcu Change 23-75 Index of Off-Nomal and Emergency Procedures 2h-75 C10.4.1, C10.4.2 Change Voltage Points 25-75 B10.0 - BlO.3 3 Manual Use of Personnel and Equipment Locks 27-75 Bh.O Add New Section - B5 5 5 5 28 75 B1.1.10 5 Change MAPIEGR Constant

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29-75 B30.2 3 c.ntrol Rod Drive System - Add Two Rod Shutdown Margin t

Limitatic.a 30-75 Bl.2.2.1 & B29 2.2 Add Use of Reactor Vessel Heatup Curves 31-75 B10 3 Added - Use of Escape, Equipment, and Personnel Locks 32-75 Bh.3 3 Delete This Section for Sodium Pentaborate Testing 33-75 D2.lk Emergency Procedure - Loss of Emergency Diesel Generator 35-75 Tab Dl.4 Annunciator 51 - Revised 36-75 B12.0 Delete in Entirety 38-75 Bil.0 Revised Special Operating Procedure Numbers 39-75 Renumbered Parts of Old "D" Section in Procedures Manual to "F" Section Numbers h0-75 Renumbered Off-Nomal Procedure D2 39 to D2.15 41-75 Rewrite of B-9 Procedure to SOP 9 Fuel Pool System I

42-75 B13 1.1, B13 2.10 Add Technical Specification Language to Operating Requirements Sections 44-75 B30.2 3 Removal of Accumulator for Maintenance 45-75 D3 1 Loss of DC Power System 47-75 C.ll Deleted From Procedures Manual l

48-75 Rewrite Reactor Operations Section to SOP-1 Section l

49-75 B5 0 Emergency Operation of Shutdown Cooling System

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i Change No.

Section and System Change 50-75 D2.18.4 Add " Manual Separation From System" 51-75 D2 23 Malfunction of Pressure Control System - New 52-75 D3 9 New Procedure - Emergency 53-75 D3 8 New Procedure - Emergency 54-75 D3 7 New Procedure - Emergency 55-75 D2 5 New Off-Nomal Procedure 56-75 D2 13 New Off-Nomal Procedure 57-75 D2 9.B Feyision and Addition of ATWAS Procedure 58-75 D3 4 New Emergency Procedure - Fuel Handling Accidents 59-75 D2.25 New Off-Nomal Procedure - Emergency Shutdown D.

Results of Surveillance Tests and Inspections Required by Technical i

Specifications The following listings show the system tested, the required test frequency, the dates tested during this report period, and the results of the test.

1.

Containment Isolation a.

System - Containment isolation valve controls and instrumentation.

Required Frequency - Quarterly (conducted monthly).

Test Dates - T30 'l was perfomed on July 1, August 1, August 29, September 30, October 28, November 26, 1975 The plant was shut down on December 6, 1975 Pre-startup check sheet C-3 (similar to T30-01) was run on this date (12/6/75) and the plant was returned to service on December 8,1975 Test T30-01 was again perfomed on December 23, 1975 Results - The automatic controls and instrumentation for eight of nine isolation valves were checked and found to function properly.

One valve (main steam drain valve M0/7065) is maintained in the closed position, de-energized and not used; therefore, testing the automatic controls of this valve is not required.

b.

System - Isolation valve leck and operability test.

Required Frequency - Twelve months (or less).

Test Dates - Test T365-04 was not required during this report

period, f

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c.

System - Containment sphere penetration inspection (visual).

Required Frequency - Twelve months (or less).

Test Dates - This test was not required during this report period.

d, System - Containment sphere integrated leak rate test.

Required Frequency - Every two years.

Test Date - This test was not required during this report period.

e.

System - Containment component leak rate test.

Required Frequency - Six months (or less).

Test Dates - Test T180-01 was perfomed on September 22 through September 27, 1975 Results - The results of the testing indicated a total leak rate which was about 46% of the lo CFR 50, Appendix J acceptance criterion which was less than 27% of the Technical Specification limit.

2.

Control Rod Drive System and Associated Tests System - Reactor safety system scram circuits (not requiring plant a.

shutdown to test).

Reguired Frequency - one month (or less).

Test Daten - T30-ol was perfomed on July 1, August 1, August 29, September 30, october 28, and November 26, 1975 The plant was removed from service on December 6,1975, check sheet C-3 (similar in content to T30-ol) was again perfomed on December 23, 1975 Results - All tests perfomed were sa isfastory with one minor exception. The test of August 1

~ 75, resulted in a full trip 4

from a single high flux level test input, traced to dirty contacts.

Subsequent testing cleared problem. This did not effect safety-related application.

b.

System - Control rod perfomance - run.

Required Frequency - Each major refueling and at least once every six months during power operation.

Test Dates - December 7, 1975 Results - The "as found" withdrawal time for fourteen control rod drives were found to be less than the 23-second minimum withdrawal time as specified in the current Technical Specifications.

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A request for a Technical Specifications change has been submitted to the NBC for consideration to remove this minimum withdrawal time requirement. The control rod drive continuous withdrawal and insertion test, including withdrawal timing, was perfomed for each CRD during the test, except for CRD E h which is valved out (selector) in the fully inserted position. Results of final timing showed all CRDs (except E 4) to be operating satisfactorily with all withdrawal times between 36 and 38 seconds.

c.

System - Control rod performance - jog.

Required Frequency - Each major refueling and at least every six months during power operation.

Test Dates - December 7, 1975 Results - Satisfactory latching and unlatching of all CRDs.

d.

System - Control rod performance - scram.

Required Frequency - Each major refueling and at least once every six months during power operation.

Test Dates - December 7, 1975 Results - The CRD scram test was performed for each CBD. The test included time from system trip to 100% of insertion. The results of the tests were all within the Technical Specificationa of ( 2 5 seconds for 90% of travel. The longest time recorded was 1 73 seconds.

System - Reactor safety systems scram circuits (requiring plant e.

shutdown.)

Required Frequency - During each major refueling outage but not less frequently than once every twelve months.

Test Dates - Not required during this report period.

f.

System - Reactor safety system response time (requiring plant shutdown).

Required Frequency - During each major refueling shutdown, but not less frequently than once every twelve months.

Test Dates - Not required during this report period.

g.

System - Control rod withdrawal permissive interlocks function.

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i Required Frequency - Twelve months or less. The refueling inter-i locks will be tested prior to each ma,1or refueling.

Test Dates - Test TR-02 was not required during this report period.

h.

System - Control rod drive friction test.

Required Frequency - During each major refueling, but not less than once each year.

Test Date - This test' was not required during this report period.

3 Emergency Cooling a.

System - Core spray system check valves.

Required Frequency - Twelve months or less.

Test Date - This test was not required during this report period.

b.

System - Post-Incident spray system automatic control operation.

Required Frequency - Twelve months or less.

Test Dates - This test was not required during this report period.

c.

System ' Reactor emergency core cooling system trip circuit.

Required Frequency - Twelve months or less.

T,est Dates - This test was not required during this report period.

d.

System - Containment sphere isolation trip circuits.

Required Frequency - During each major refueling shutdown, but not less frequently than once every twelve months.

Test Dates - This test was not required during this report e.

System - Emergency condenser outlet valves test.

Required Frequency - Twelve months or less.

Test Dates - This test was not required during this report period.

f.

System - High energy pipin6 leakage inspection.

Required Frequency - Monthly when turbine generator is in service.

Test Dates - July 15, August 12, Septedber 8, October T, Novedber 4 and Decedber 2, 1975 Results - Satisfactory, no leakage detected.

g.

System - Primary system leakage test.

Required Frequency - Daily.

Test Dates - The calculation was made daily from July 1 through December 31, 1975, with the exception of Decedber 7, 1975 when the plant was shut down. This test is only required when the I'

system is in the " hot" condition.

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Results - Satisfactory, 1 gpm unidentified.

4.

Miscellaneous Systems a.

System - Reactor shutdown margin test.

Required Frequency After each refueling, after certain core component changes, if the system is cooled to atmospheric con-ditions and after 35,000 MWdt.

Test Date - This test was not required during this reporting period.

b.

System - Nil Ductility Transition Temperature (NDTT) calculation.

Required Frequency - At least once per year.

Test Date - This calculation was not required during this reporting l

period.

c.

System - Moderator temperature coefficient test.

Required Frequency - Followit, each major refueling outage.

Test Date - This test was not required during this reporting period.

d.

System ~ Suberiticality checks Required Frequency

'>uring core alterations which increase core reactivity.

Test Date - This test vos not required during this reporting period.

1 System - In-service primary system inspection.

e.

Required Frequency - A continuing program is being conducted during some major refueling outages.

Test Date - This test was not required during this reporting period.

f.

System - Refueling operation controls.

Required Frequency - Each major refueling.

Test Dates - This test was not required during this report period.

g.

System - Reactor refueling safety system sensors and trip devices.

Required Frequency - Each major refueling.

Test Dates - This test was not required during this report period.

h.

System - Recirculation pump valve interlock test.

Required Frequency - Twelve months.

Test Date - This test was not required during this report period.

5 Poison System a.

System - Liquid poison system firing circuit test.

Reqttired Frequency - Two months or less.

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Test Dates - Test T60-01 was performed on August 27, October 28, and December 23, 1975 Results - Satisfactory.

b.

System - Explosive valve from equalizing line.

Required Frequency - Twelve months or less.

Test Date - This test was not required during this report period.

c.

System - Explonive valve from nonequalizing lines.

Required F'requency - Twelve months or less.

j Test Date - Test T365-12 was not required during this report period.

I 6.

Radiation Monitoring a.

System - Air ejector and off-gas monitoring system.

Required Frequency - One month or less.

Test Dates - July 24. August 28. Septesber 18, October 28, Novesber 21 and December 16, 1975 Results a Checks showed the calibration to be satisfactory. The automatic closure function of the isolation valve timer was checked and showed the timer calibration to be satisfactory (within 3%

of the maximum timer setting) and the isolation valve closed as specified.

b.

System - Calibration and functional test of the stack gas monitoring system.

Required Frequency - One month or less.

Test Dates - July 30, August 28, Septesber 18, October 28 and 31, November 24 and Decesber 16, 1975 Results - The stack gas monitoring system was checked using the built-in Cs-137 calibration source. The instrument check showed the calibration to be satisfactory, resulting in the alarm oc-curring within the specified 0.1 curie per second release rate.

An additional calibration of stack gas monitoring system is a comparative calibration used to demonstrate operations of the monitor and to detect gross calibration changes and/or instrument drift. All calibrations were within the acceptance criteria of 1 0% since recalibration of the monitor with standard liquid sources.

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c.

System - Analyses of stack gas particulate and iodine filters.

Required Frequency - Weekly.

Test Dates The analyses were conducted weekly.

Results - The results of analyses of the stack gas particulate filter and iodine filter are reported in tems of curies released in Appendix A of this report.

d.

System - Calibration of emergency condenser vent monitor.

Required Frequency - One month or less.

Test Dates - July 28, August 28, September 18, October 29, November 20 and December 20, 1975 l

Results - The emergency condenser vent monitors are checked by comparing with a calibrated portable instrument. The checks i

showed the ent monitor calibrations to be satisfactory with all

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monitor checks within 10% of full scale. Alam points were foundto'belessthan10mR/hplusba'ckground.

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e.

System - Calibrac, ion of canal liquid process monitor.

Required Frequency - One month or less.

Test Dates - July 24, August 28, September 18, October 28, November 20 and December 16, 1975 Results - The calibration of the canal liquid process monitor is a comparatis calibration used to demonstrate operations of the moniter and to detect gross calibration change and/or instrument drf't.

All calibrations were within the acceptance criteria of 1 0f, since recalibration of the monitor with standard liquid 3

sources.

f.

Syytem - Canal liquid collection sample.

Required Frequency - Daily.

Test Dates - The analyses was conducted daily.

Results - Satisfactory.

g.

System - Calibration of area monitors.

Required Frequency - One month or less.

Test Dates - July 24, August 28, September 18, October 28, November 20 and December 16, 1975 Results - The area monitor calibrations are checked by comparing f

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i readings with a calibrated portable instrument. The checks showed the area monitor calibration to be satisfactory with most monitors within 10% and all monitor calibrations within

!20%.

h.

System - Calibration of all liquid process monitors (except the canal monitor).

Required Frequency - Three months or less.

Test Dates - July 24, August 28, September 18, October 28, November 21 and December 16, 1975 Results - The calibration of the liquid process monitors (except the canal monitor which is reported separately) is a comparative calibration used to demonstrate operation of the monitor and to detect gross calibration changes and/or instrument drift. All calibrations were within the acceptance criteria of 1 0%.

3 E.

The Result of Any Periodic Contairnent Le'ak Rate Test Perfomed During the Report Period The biannual containment integrated leak rate test was not required during this report period.

F.

Technical Specifications Changes During this period, Technical Specification changes were authorized by the Commission as follows:

1.

Change #46, Amendment 9 to the license which incorporates revised administrative controls to meet the regulatory positions of Guides 1.8, 1 33, 1.16.

G.

Changes in Plant Operating Organization Involving Key Supervisory Personnel None H.

Refueling Infomation 1.

Name of Facility: Big Rock Point 2.

Scheduled Date for Next Refueling Shutdown: January 30, 1976 3

Scheduled Date for Restart Following Refueling: May 15, 1976 4.

Resumption of Operation Requiring Technical Specification Change or License Amendment:

Licensing of Gl-U fuel Licensing of Reactor Depressurization System

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Unit On Line: December 8, 1975 - 0516 hours0.00597 days 0.143 hours 8.531746e-4 weeks 1.96338e-4 months Length of Outage: 49 hours5.671296e-4 days 0.0136 hours 8.101852e-5 weeks 1.86445e-5 months , 43 minutes Discussion: A weekly pipe tunnel inspection on December 2,1975 j

detected a considerable amount of steam from under the front end of the turbine. Attempts to install a temporary patch on the pipe were not successful, and, therefore, the unit was taken off the line.

The leak was a hole at the welded. junction of the high-pressure turbine casing reducer and the 1/2" drain pipe, from the No. 2 stage drain header. A temporary patch was made and the unit was returned to service. Plans are to inspect the remaining drains during the 1976 refueling outage and replace the patched drain and others that are found to be bad.

V.

SAFETY-RELATED MAINTENANCE NOTE: Dates contained in this section generally refer to the weekly period when the mainterfance was perfomed.

A.

Reactor Protection and Control System Instrumentation 1.

Neticron Monitoring Channel No.1

?.nere was no major corrective maintenance perfomed on this channel i

during the reporting period.

2.

Neutron Monitoring channel No. 2 11/20/75-Component: Power Channel No. 2 Picoammeter a.

Cause of Malfunction: Suspect spurious trip and erratic operation at lo= c. ar levels.

Effect on Safe Operation: None - suspect failure resulted in trip and reactor scram while testing Channel No.1.

Technical Specifications and plant design allow for rencval of one neutron l

monitoring channel for maintenance.

Corrective Action: Following replacement of picoammeter, faulty operation of the range switch was noted when switching from the 125% power position to the " Test Trip" position. The range switch was replaced to correct this discrepancy.

I Following switch replacement, erratic operation was noted in the more sensitive positions (125 x lO~I % power) of the range switch.

(

17 gi

l Subsequent trouble-shooting resulted in replacement of the dual high voltage power supply (did not correct the problem) and re-placement of chamber coaxial cables from the chamber to the chamber drive head.

Reference Maintenance Orders 75-NM-1061 75-NMS-1062 75-NMS-1063 75-NMS-10614 Precautions to Provide for Reactor Safety During Repair: Work perforined on this channel was performed under procedural guidance and/or trip insertion while being repaired.

b.

11/20/75-Component: Power Channel No. 2 Dual High Voltage Power Supply.

Cause of Malfunction: Defective negative supply causing loss of input AC fuse.

Effect on Safe Operation: None - Failure of negative power supply places channel in " ready" condition for single upscale trip.

Technical Specifications and plant design allow for removal of one neutron monitoring channel for maintenance.

Corrective Action: Replaced the dual high voltage power supply with a spare unit. Failures of this type are within the design limitations of the equipment. Ref: MO 75-NMS-1069 Precautions to Provide for Reactor Safety During Repair: The neutron monitoring channel remained in the downscale trip condition during power supply replacement.

12/11/75-component: Neutron Monitoring Channel No. 2 Piconmmeter c.

very erratic at low power levels (shutdown condition).

Cause of Malfunction: Faulty negative high voltage power supply.

Effect on Safe Operation: None - Reactor ir.,hutdown condition; other two picoammeters reading on scale between the upscale (120%)

and downscale (5%) trip points and inserted a high flux trip on No. 2 picosmmeter.

(

18

1 5

Reload Gl-U - October 13, 1975 and February 9, 1976 Reactor Depressurization System - March 10, 1975 6.

Important Licensing Considerations:

Gl-U Fuel Licensing Licensing of Reactor Depressurization System III. P_0WER GENERATION Report Period Total to Date A.

Themal Power Generated (MWht) 808,734 14,296,622 B.

Gross Electric Power Generated (MWhe) 25b,400 4,549,196 C.

Net Electric Power Generated (MWhe) 240,333 9 4,308,143 9 D.

Hours Critical 4,335 9 80,146 5 E.

Hours Generator on Line 4,316.6 78,281.1 F.

Maximum Dependable Capacity (MWe net) 71 71 G.

Reserve Shut,down Hours 0

0 IV.

SHUTDOWNS A.

fype - Forced Unit Off Line: October 30, 1975 - 0407 hours0.00471 days 0.113 hours 6.729497e-4 weeks 1.548635e-4 months .

Unit On Line: October 30,1975 - 0948 hours0.011 days 0.263 hours 0.00157 weeks 3.60714e-4 months .

Length of Outage: 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , 41 minutes.

Discussion: A leak in an elbcu on the turbine stage drain line in the pipa: tunnel was discovered earlier during this period. The leak deteriorated until the unit was brought off the line. The leak was then repaired and the unit was returned to service. This outage also coincided with some repair work on the Initial Pressure Regulator (IPR).

B.

Type - Scheduled Unit Off Line: November 13, 1975 - 0452 hours0.00523 days 0.126 hours 7.473545e-4 weeks 1.71986e-4 months .

Unit On Line: November 15,1975 - 0155 hour0.00179 days 0.0431 hours 2.562831e-4 weeks 5.89775e-5 months s-Length of Outage: 45 hours5.208333e-4 days 0.0125 hours 7.440476e-5 weeks 1.71225e-5 months , 7 minutes.

Discussion: The shutdown occurred as a result of numerous condenser leaks. Leaks have been increasing to a point where the condensate demineralizer resins were being used up at a high rate, and new resins were hard to obtain due to a nationwide shortage of supply. As a result three (3) condenser tubes were plugged an.1 the unit was returned to service.

C.

Type - Scheduled Unit Off Line: December 6,1975 - 0333 hours0.00385 days 0.0925 hours 5.505952e-4 weeks 1.267065e-4 months 16

t.

Corrective Action: Replaced the dual high voltage power supply to Channel No. 2 picommmeter. Bench repair of the defective supply consisted of replacement of one electron tube and an open fine voltage adjust control in the positive supply. Tailures of this type are within the design limitations of the equipment.

Precautions to Provide for Reactor Safety During Repair: Reactor in shutdown condition, other two picoammeters on scale between the upscale (120%) and downscale (5%) trip points, and a high flux trip inserted into Channel No. 2.

High voltage power supply was replaced under the guidance of Procedure No. INMS-13 (MO 75-NMS-1071).

3 Neutron Monitoring Channel No. 3 8/7/75-Component: Static Power Inverter (normally supplies power a.

to Channel No. 3 picoammeter) tripped off with station battery on overcharge.

Cause of Malfunction: Suspect drifting high voltage trip point setting on inverter and/or high battery voltage.

Effect on Safe Operation: None - Channel No. 3 is supplied from the alternate power supply, Instrument and Control Bus No. 3, when the static inverter is not available. Failure of the inverter initiates a Channel No. 3 trip.

Cogyactive Action: The voltage was found to be two (2) volts high on the station battery charger. This was adjusted in accordance 4

with the vendor's instructions and calibration nf the panel volt-meter was also checked.

(MO75-SPS-loo 3)

Precautions to Provide for Reactor Safety During Repair: During the equalizin6 charge period, Channel No. 3 was powered from the alternate power supply.

b.

10/9/75 - Component: Static Power Inverter (normally supplies power to Channel No. 3 picommmeter). Inverter output voltage dropped to 102V a-c from a normal of llW a-c (Station battery on over-charge).

Cause of Malfunction: Suspect, from a previous similar condition when station battery was on over-charge, that the inverter high voltage trip point setting was causing this problem.

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19

Effect on Safe Operation: None. Channel No. 3 is supplied power from the alternate power supply, Instrument & Control Bus 3, when the static inverter is not available.

Corrective Action: The static inverter was replaced with a cali-brated operational spare unit. The failed unit was sent to Labora-tory Services for testing of the high voltage trip point.

(Mo75-M4s-loS5)

Precautions to Provide for Reactor Safety During Repair: The static inverter was replaced under the guidance of Procedure No. IID4S-7.

11/6/75 - Component: Static Power Inverter S/N-004 (normally c.

supplies power to Channel No. 3 picoammeter) tripped off when station battery was put on over-charge.

Cause of Malfunction: Suspect from previous similar condition with static inverter S/N-001, that high voltage trip setting has drifted downward to coincide with the station battery equalizing charge.

Effect on Safe Operation: None. Channel No. 3 is supplied power from the alternate power supply, Instrument & Control Bus 3, when the static inverter is not available.

Corrective Action: The static inverter was replaced with a recently calibrated operational spare unit. The failed unit will be sent to Laboratory Services for testing of the high voltage trip setting.

(Mo 75-NMS-los8)

Precautions to Provide for Reactor Safety During Repair: The static inverter was r6placed under the guidance of Procedure No. IIO4S-T.

4.

Neutron Monitoring Channel No. 4 7/17/75-Component: Channel No. 4 Log-N/ Period Amplifier.

a.

Cause of Melfunction: Instability in the period indication.

Effect on Safe Operation: None. The intemediate neutron monitor-ing channels are not required above 5% reactor power.

Correcti_ve Action: Replaced a noisy " Infinity Set and Trip Test" potentiometer.

Precautions to Provide for Reactor Safety During Repair: The Log-N/

Period amplifier was replaced under the guidance of Procedure INMS-2 (MO 75-NMS-lo39).

?

b.

10/2/75 - Component: Intemediate Channel No. 4 Log-N/ Period Amplifier Cause of Malfunction: Per$M amplifier tube defective resulting in short period alams.

Effect on Safe Operation: None. The Technical Specifications and plant design do not require this instrument to be in sewice when reactor power is above 5% rated power.

Corrective Acticn: Replaced the defective unit with the opera-tional spare; bench repair subsequently perfomed on the defective unit. Failures of this type are within the design limitations of the equipment.

(M0 75-104S-1054)

Precautions to Provide for Reactor Lafety During Repair: Replace-ment perfomed under guidance of Procedure I104-2.

12/11/75 - Component: Neutron Monitcring Channel No. 4 was erratic c.

at low power levels (~ 4 x 10-5) percent power).

Cause of Malfunction: Suspected Log-N amplifier.

Effect on Safe Operation: None - Reactor was in shutdown con-dition below 4% power on picoammeters and a full scram condition was initiated when removing the Log-N amplifier.

Corrective Action: Channel No. 4 Log-N amplifier was replaced with an operational spare unit. Subsequent trouble shooting traced the problem to the chamber polarizing high voltage power supply.

Precautions to Provide for Reactor Safety During Repair: Reactor in shutdown below 4% poser and a full scram condition was initiated while replacing the Log-N amplifier. Unit replaced under the guidance of Procedure INMS-2 (MO 75-NMS-1070).

d.

12/11/75-Component: Neutron Monitoring Channel No. 4 erratic at low power levels (* 4 x 10~E) percent power.

Cause of Malfunction: Faulty positive high voltage supply.

Effcet on Safe Operation: None - Reactor in shutdown condition below 4% power on picoammeters and a full scram condition initiated when high voltage power supply is removed.

Corrective Action: Replaced the dual high voltage power supply to Channel Log-N amplifier. Subsequent repair to the failed 21 w

supply resulted in replacement of tnree marginal electron tubes and an open fine voltage adjust control in the positive supply.

Failures of this type are within the design limitations of the equipment.

Precautior.s to Provide for Reactor Safety During Repair: Reactor in shutdown condition and belov 4% power with a full scram condi-tion in. The high voltage power supply to Channel N.4 was re-placed under the guidance of Procedure INMS-14 (MO 75-NMS-1072).

5 Neutron Monitoring Channel No. 5 9/25/75 - Component: Intemediate Channel No. 5 Log-N/ Period a.

Amplifier.

Cause of Malfunction: Internal 150V d-c power supply capacitor failure.

Effect on Safe Operation: None. The Technical Specifications and plant desig: do not require this ' instrument u be in service when reactor power is above 5% rated power.

Corrective Action: Replaced the defective unit with the operational spare. Bench repaired and calibrated. (MO 75-NMS-1052)

Precautions to Provide for Reactor Safety During Repair: Replace-ment perfomed under guidance of Procedure INM-2.

6.

Neutron Monitoring Channel No. 6 There was nu major corrective maintenance perfomed on this channel during the reporting period.

7 Neutron Monitoring Channel No. 7 10/2/75 - Component: Start-up Channel No. 7 Log Countrate Meter a.

Cause of Malfunction: Gassy period amplifier tube resulting in short period alams.

Effect on Safe Operation: None. The Technical Specifications do not require this instrument to be in service during power operation.

Corrective Action: Replaced the defective unit with the opera-tional spare; bench repair subsequently perfomed on the defective unit. Failures of this type are within the design limitations of the equipment.

(MO75-NMS-1056)

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22

Precautions to Provide for Reactor Safety During Repair: Replace-ment perfomed under guidance of Procedure INM-3 8.

Reactor Protection System 7/24/75 - Component: No. 2 Reactor Protection M-G Set a.

Cause of Malfunction: Inboard flywheel bearings were noted to be running noisy and warmer than usual.

Effect er. Safe Operation: None.

Corrective Action: Replaced bearin6s and aligned flywheel motor generator couplings.

Precautions to Provide for Reactor Safety i aring Repair: Work was performed under Procedure MRPS-1 with No.1 M-G set in service, b.

8/7/75 - Component: No.1 Rod position M-G Set Cause of Malfunction: Addition of grease fittings to allow proper lubrication of inner and outer race on flywheel bearings.

Effect on Safe Operation: None.

Corrective Action: Added grease fittings to No.1 M-G set Precautions to Provide for Reactor Safety During Repaie: Not applicable, M-G set not out of service.

8/7/75 - Component: No.1 M-G Set.

c.

Cause of Malfunction: Vibration and rough running reported.

Inspection to check alignment of motor and generator drive shaft.

Effect on Safe Operation: None Corrective Ac+1on: None required at this time. Bearing temperature and vibration were acceptable.

Precautions to Provide for Reactor Safety During Repair: Not applicable.

d.

9/11/75 - Component: Scram Dump Tank Level Switch RD08-B Cause of Malfunction: 1S-RD08-B failed to reset properly on

. Tune 6, 1975 Reset was assisted manually (by tapping of switch case).

Effect on Safe Operation: None. Failure of a switch of this type to reset properly normally indicates that the switch will have a more positive trip action.

i 23

Corrective Action. T~pection of the switch indicated no damage or misoperation. The mechanism had freedom of operation and exhid Md good reset characteristics. (MO75-RPS-1010)

Precautions to Provide for Reactor Safety During Repair: The circuit breaker for this sensor was placed in the trip position to initiate a single channel reactor protection system trip during the time the sensor was inspected.

10/9/75 - Component: No.1 Reactor Protection System Motor -

e.

Generator (output voltage graduale increasing from norinal 120 I 2V a-c to 123V a-c).

Cause of Malfunction: Defective (aged) voltage regulator electron tube in the voltage control unit.

Effect on Safe Operation: None. The motor-generator unit is equipped with an over-voltage relay to trip the output when 1h0V a-c.is exceeded. Loss of generator output voltage pla::es the protection system in a trip (failsafe) condition.

Corrective Action: The defective electron tube was replaced and the output voltage returned to 119V a-c.

Failures of this type are within the design 11midtic..:; of the equipment. (MO75-RPS-1011).

Precautions to Provide "or Reactor Safety During Repair: The alternate power supply (Instrument & Control Bus No. 3) was used to supp'y Bus No. 2 during repair.

f.

10/30/75-Component: Reactor Protection Channel No. 2 High Reactor Pressure Switch RE07B failed in the trip condition and would not reset.

Cause of Malfunction: Dust or dirt ap7arently holding contacts open.

Effect on Safe Operation: None. Reactor Protection Channel No. 2 was in the scram condition.

Corrective f.ction: Emergency situation - No adjustments made.

Contact assembly was rotated to wipe contacts. Positive relay operation was indicated. (MO75-RPS-1015)

Precautions to Provide for Reactor Safety During Repair: Reactor Protection Channel No. 2 was in the scram condition.

(

2h

I B.

Radioactive Effluent Monitoring Systems 1.

Air Ejector Off-Gas System 8/28/75-Component: Log Radiation Monitors a.

Cause of Malfunction: Setpoint change (S-001) to compensate for off-gas flow increase to 11 cfm flow.

Effect on Safe Operation: Would have allowed exceeding release rates if setpoint change not perfomed.

Corrective Action: Reduced setpoints on log radiation monitor 3

3 as follows; Alam: 1.6 x 10 units to 1.0 x 10 units (30,000 pCi/sec);

3 3

Isolation Valve Closure: 2 7 x 10 units to 1 5 x 10 units (50,000pCi/sec).

Precautions to Provide for Reactor Safety During Repair: Two units provide redundancy in this application.

b.

8/28/75 - Component: Off-Gas Monitor Flow Meter

~

Cause of Malfunction: Air-in-leakage at the flow-meter around the glass tube ends.

Effect on Safe Operation: None.

Corrective Action: Air-in-leakage was located and temporarily repaired using Glyptol. Pemanent repairs vill be perfomed upon fomulation and approval of appropriate pla'it procedures.

Precautions to Provide for Reactor Safety During. Repair: No procedure was required for this work. Back-ups for this system during repair were provided by the stack gas system and constant radiation protection surveillance.

(75-WGS-239-06) 9/4/75 - Component: No.1 Log Radiation Monitor c.

Cause of Malfunction: Trip circuit failed to trip off-gas isolation valveat50,000pCi/sec(15x10' units).

Effect on Safe Operation: The automatic trip nomally would be relied on to clJse the off-gas valve, however, administratively this valve is not relied on to operate in this mode due to previous deficiencies in the off-gas system related to the integrity of the off-gas line. In this instance, no undue hazard was deemed to exist.

(

25

Corrective Action; The relay units for the alam and trip circuits i

were swapped (and returned) to detemine cause of the failure. The problem corrected itself and has not recurred (the units are tested daily for alam and trip function).

Precautions to Provide for Reactor Safety During Repair: Two units provide redundancy in this application.

d.

10/2/75-Component: Channel No.1 Off-Gas Log Radiation Monitor.

Cause of Malfunction: Unknown. Investigation of previous failun (seepreviousitemof9/4/75).

Effect on Safe Operation: None. Procedure No. IWGS-2 precautions insure that the off-gas timer monitor selector switch is not selecting the unit (Log Radiation Monitor) that will be removed.

Corrective Action: Checked all circuits in No.1 Log Radiation Monitor and the unit was calibrated according to Procedure No.

TR h0 " Electronic Calibration of Off-Gas Instrumentation" (Log Radiation Monitor). (Mo 75-MGS-loo 7)

Precautions to Provide for Reactor Safety During Repair: Pro-cedure IWGS-2 was used to remove and install the No.1 Log Radia-tion Monitor.

11/6/75 - Cceponent: Off-gas System Purge Air Solenoid RL-25 e.

was sticking and diluting sample to detector.

Cause of Malfunction: Normal wear.

Effect on Safe Operation: None, instrumentation and off-gas by-par,s line in service.

Corrective Action: Replaced the internal disc disc holder and disc spring, body gasket, core spring and core assembly body gasket.

Precautions to Provide for Reactor Safety During Repair: Followed work Procedure 75-WGS-230-03 f.

11/20/75-Component: No.1 Log Radiation Monitos high trip failed (actuates the off-gas timer).

Cause of Malfunction: Relay in c,he high trip circuit failed to de-energize (de-energizing starts the off-gas timer if unit is selected for control).

t Effect on Safe Operation: None. The No. 2 Log Radiation Monitor 26

I was selected for control and No.1 Log Radiation Monitor was caution tagged for surveillance and not to be used for control.

Corrective Action: The unit was removed (while in the failed condition) and components checked in the high trip circuit ahead of the relay. Two transistors and a capacitor were replaced in the circuit. (Mo75-woS-lolo)

Precautions to Provide for Reactor Safety During Repair: The No. 2 Log Radiation Monitor was used for control of the off-gas timer and the No.1 Log Radiation Monitor was removed and installed

'nder the guidance of Procedure IWGS-2. Bench calibration Proce-dure TR-20 was also performed on the unit before reinstallation.

2.

Stack Gas Radiation Monitoring 8/7/75 - Component: Erratic Count Rate on both the gross and a.

single isotope channels.

Cause of^ Malfunction: Defective scintillation detector.

Effect on Safe Operation: None. Backup monitoring of the stack gas is provided by the air-ejector off-gas monitors and by in-stalled samplin6 devices.

Corrective Action: Replaced the defective detector with a spare unit and checked calibration.

Precautions to Provide for Reactor Safety During 'epair: The air ejector off-gas monitors provide backup for this monitoring system. Removal of this system from service is permitted by the Technical Specifications provided repairs are promptly made and the system returned to ee vice. (MO 75-SGM-LOOT) 3 Liquid P:-ocess Monitoring 8/14/75 - Component: Radvaste Liquid Process Monitor loss of a.

indication.

Cause of Malfunction: Open viring on the scintillation detector connector assembly.

Effect on Safe Operation: None. The process channel was not in service at the time of failure. Failure occurred during the calibration process from flexing of the connector assembly.

(

27

-l

4 Cer.~er.tive Action: Repaired the connector wiring and checked calibratf on of the channel with the test source.

Precautions to Provide for Reactor Safety During Repair: None.

The channel was not in service at the time of failure.

b.

8/21/75 - Component: Radwaste to canal linear count rate meter.

Cause of Malfunction: Electron tube failure.

Effect on Safe Operation: None. Failures of this type are within the design limitations of the equipment. Removal of this system from service is permitted by the Technical Specifications provided repairs are promptly made and the system returned to service.

Corrective Action: Electron tube replacement and bench calibration following unit replacement with spare.

Precautions to Provide for Reactor Safety During Repair: Not applicable. (Mo 75-LPM-1010) 8/21/75 - Component: Reactor cooling water linear count rate c.

meter.

Cause of Malfunction: Electro' tube failure.

Effect on Safe Operation: No ne. Failures of this type are within the design limitations of the equipment. Removal of this system from service is permitted by the Technical Specifications provided repairs are promptly made and the system returned to service.

Corrective Action: Electron tube' replacement and bench calibration following unit replacement with spare.

Precautions to Provide for Reactor Safety During Repair: Not applicable. (MO75-LPM-1012)

C.

Containment Sphere Isolation System 1.

8/28/75-Component: Supplyventvalve,CV/4097, operator-to-valve flange.

Cause of Malfunction: Check for tightness of bolts to verify torque of 100 1 5ft/lbs.

Effect on Safe Operation: None.

Corrective Action: Bolts were checked and verified to be properly torqued.

Precautions to Provide for Reactor Safety During Repair: Repairs I

were within the skills of the repair man who was given specific 28

i instructions on the maintenance order not to loosen the bolts.

(75-CIS-238 -06) 2.

9/11/75 - Component: Personnel lock Cause of Malfunction: Failure of outer door to completely close occasionally.

Effect on Safe Operation: None, centainment isolation was not affected and the personnel lock was operable during repairs Corrective Action: Adjusted miervsvitch on the outer door and screwed one cylinder back into its yoke.

Precautions to Provide for Reactor Safety During Repair: Not appl $ cable.

3 9/25/ '5 - Component: Personnel lock Cause of Malfunction: Improper closing of the personnel lock outer door.

Effect on Safe Operation: None, containment isolation was not affected and the personnel lock was operable during repair.

Corrective Action: Adjusted the outer door limit switch.

Precautions to Provide for Reactor Safety During Repair: Not applicable.

4.

10/2/75 - Component: Personnel lock inner door did not close in timely fashion without assistance.

Cause of Malfunction: Lov flow on closing flow control valve.

Effect on Safe Operation: None, the lock was operable during repair.

Corrective Action: Decreased the closing time by increasing the flow through the closing flow control valve. The door was subsequently adjusted.

Precautions to Provide for Reactor Safety During Repair: Not applicable, containment isolation was unaffected.

D.

Energency Power System 1.

7/24/75 - Component: Diesel Generator Cause of Malfunction: No indication of local voltage or machine frequency during periodic test start.

Effect on Safe Operation: Not applicable.

(

29

(

l Corrective Action: Replaced fuse and remounted from horizontal to vertical position. Test ran diesel generator after maintennnce.

Precautions to Provide for Reactor Safety During Repair: Both 45 kV

s.

and 138 kV lines were available for service during re

.r 2.

8/21/75 '- Component: Emergency Diesel Generator Cause of Malfunction: Metering circuit fese failures were determined to be resulting from fatigue failure due to vibration of the fuse link.

Effect on Safe Operation: None.

Corrective Action: Renewable fuses were replaced with one-time fuses which contain a powder substance to eliminate vibration of the fuse link. Diesel generator metering was tested subsequent to repairs.

Precautions to Provide for Reactor Safety During Repair: Both the h6 kV and 138 kV lince were available during repair and the Plant Superintendent's pemission to remove diesel generator from service was obtained prior to repair.

(75-EPS-224-10) 3 9/25/75 - Component: Emergency Diesel Generator Cause of Malfunction: Investigation of clicking noise in the starter during diesel generator starting of September 22, 1975 Effect on Safe Operation: None, the diesel was not rendered inoperable.

Corrective Action: During investigation on September 23, 1975, the diesel started normally. The clicking was attributed to a timer switch controlling the starter circuit. This is normal for the timer switch.

Precautions to Provide for Reactor Safety During Repair: The work was performed with both 46 kV and 138 kV lines available. The work performed did not render the diesel inoperable.

4.

10/2/75 - Component: Emergency Diesel Generator cooling water pump Pr. eking leaked excessively.

qse of Malfunction: Loose packing nuts on cooling water pump.

c on Safe Operation: None, diesel generator was available for

. ~.

Correct* Action: Turnedpackingnuts1/2flattostopleak-off and rect.ecked during weekly start-up. Nuts were subsequently I

turned 1/2 flat during weekly start-up. Leak-off was stopped.

30

I Precautions to Provide for n'ector 0;fety During Repair: Packing adjustments on the cooling water pump did not remove the generator from service.

E.

Emergency Condenser System No work in the Emergency Condenser System was performed during this report period.

F.

Primary Coolant System 1.

7/24/74 - Component: Spare recire pump seal.

Cause of Malfunction: Excessive leakage from seal when in operation.

Effect on Safe Operation: Not applicable, seal removed from service.

Corrective Action: Disassembled, cleaned, lapped, and rebuilt seal.

Precautions to Provide for Reactor Safety buring Repair: Not applicable.

G.

Shutdown Cooling System No work in the Shutdown Cooling System was performed during this report period.

H.

Control Rod Drive System 1.

CRD Filters 10/2/75 - Component: No.1 CRD filter had high differential pressure.

a.

Cause of Malfunction: The No. 1 filter was clogged.

Effect on Safe Operation: None, only one filter is required during operation.

Corrective Action: Replaced No.1 CRD filter with spare. Cleaned the removed filter for future use.

Precautions to Provide for Reactor Safety During Repair: Work was performed per MCRD 6, Rev. 1.

The No. 2 filter was in service, b.

10/2/75 - Component: No. 2 CRD filter had high differential pressure.

Cause of Malfunction: The No. 2 filter was clogged.

Effect on Safe Operation: None, only one filter is required during operation.

Corrective Action: Replaced No. 2 CRD filter with spare. Checked for leaks with none noted. Cleaned the removed filter for future use.

1 Precautions to Provide for Reactor Safety During Repair: Work was performed per MCRD-6, Rev.1.

The No.1 filter was in service.

i 31

y l'

10/16/75-' Component: The No.1 C3D filter had high differential

j c.

pressure.

Cause of Malfunction: The filter was clogged.

Effect on Safe Operation: None, only one filter is required during

[

operation.

Corrective Action: Replaced the No. 1 filter with a spare. Checked i

for leaks and none were noted. Cleaned the removed filter for future use.

Precautions to Provide for Reactor Safety During Repair: Work was performed per MCRD-6, Rev.1. The No. 2 filter was in service.

l d.

10/23/75 - Component: The No. 2 CRD filter had high differential pressure.

1 Cause of Malfunction: The filter was clogged.

Effect on Safe Operation: None, only one filter is required during operation.

Corrective Action: Replaced the No. 2 filter with a spare.

~

Checked for leaks and none were noted. Cleaned the removed filter for future use.

Precautions to Provide for Reactor Safety During Repair: Work vac performed per Procedure MCRD 4, Rev.1.

The No. 1 filter was in service.

10/30/75-Component: The No. 2 filter had high differential pressure e.

Cause of Malfunction: The filter was clogged.

i I

Effect on Safe Operation: None, only one filter is required during power operation.

Corrective Action: Replaced the No. 2 filter with a spare and i

cleaned the removed filter for future use.

Precautions to Provide for Reactor Safety During Repair: The work was performed per MCRD 6, Rev.1 vith the No.1 filter t

in service.

f.

12/4/75 - Component: No. l' CRD filter had high differential pressure.

Cause of Malfunction: The filter was clogged.

Effect on Safe Operation: None, only one filter is required for operation.

(

32 E

i -

a.- - - -.

i Corrective Action: The filtar was : eplaced and the clogged filter cleaned for reuse.

Precautions to Provide for Reactor Safety During Repair; The No. 2 filter was in service and work was performed per MCRD-6, Rev. 2.

g.

12/11/75-Component: The No. 2 filter had high differential pressure.

Cause of Malfunction: The filter was clogged.

Effect on Safe Operation: None - only one filter is required for operation.

Corrective Action: Replaced the filter with a spare and cleaned the clogged filter for future use.

Precautions to Provide for Reactor Safety During Repair: Work was performed us'ng MCRD 6, Rev. 2 with the No.1 filter in service.'

h.

12/18/75-Component: The No.1 filter had high differential pressure.

Cause of Malfunction: The filter was clogged.

Effect on Safe Operation: None, only one filter is required to be in service during operation.

Corrective Action: Replaced the clogged filter with a spare and reconditioned the clogged filter.

Precautions to Provide for Reactor Safety During Repair: Work was performed using MCRD-6, Rev. 2 with the No. 2 filter in service.

2.

CRD Accumulators 7/17/75 - Component: Accumulator A 4.

a.

Cause of Malfunction: Gas leak through spacer ring beneath locknut.

Effect on Safe Operation: No effect, one rod driven may be removed from service.

Corrective Action: Installed new "O" rings and backup rings.

Precautions to Frovide for Reactor Safety During Repair: Work j

performed using Procedure MCRD-3

('

33

b.

7/17/75 - Component,: Accumulator D-6.

Cause of Malfunction: Gas leaking through jamnut.

Effect on Safe Operation: No effect, one rod drive may be removed from service.

Corrective Action: Installed new "O" rings and backup rings.

Precautione to Provide for Reactor Safety During Repair: Work perfonned using Procedure MCRD 7/29/75 - Component: Accumulator D 1 c.

Cause of Malfunction: Nitrogen gas :c aking through reducer on gas side piping to leak detection instrumentation.

Effect on Safe Operation: Not appl $ cable.

Corrective Action: Mone taken, no leak was noted after inspecting with liquid leak check.

Precautions to Provide for Reactor Safety During Repair:

Procedur'e MCRD-3 used.

d.

8/28/75-Component: F 4 Accumulator.

Cause of Malfunction: Nitrogen p.as leaking from the spacer ring between the jamnat and backup ring on the bottom accumulator.

Effect on Safe Operation: None.

Corrective Action: New seals were installed and a leak test was performed after installation.

Precautions to Provide for Reactor Safety During Repair: Work was controlled under Procedure MCRD-3, Rev. 4 with only one ac-cumulator rendered inoperable. (75CRD-206-08) 8/28/75-Component: E 4 Accumulator.

e.

Cause of Malfunction: Nitrogen gas leaking from between the spacer and locknut.

Effect on Safe Operation: None.

Corr.ctive Action: New seals were installed ana a leak test was performed after installation.

Precautions to Provide for Reactor Safety During Repair: Work was controlled under Procedure MCRD-3, Rev. 4 with only one accumulator rendered inoperable. (75 CRD-224-03)

(

34 4

v-

-w

f.

9/18/75 - Component: Accumulator C-1.

Cause of Malfunction: Leakage from plug at bottom of accumulator.

Effect on Safe Operation: None, one accumulator can be rendered inoperable.

Corrective Action: A new burst disc assembly was installed.

Precautions to Provide for Reactor Safety During Repair: Work was controlled per Procedure MCRD-3, Rev. 4 with only one accumulator rendered inoperable.

g.

11/20/75 - Component: The F 4 accuculator had " low pressure" and " leaks" alarms.

Cause of Malfunction: Normal wear on "O" ring and bladder.

Effect on Safe Operation: None, the plant uns shut down with all control rods inserted.

Corrective Action: The accumulator bladder and "O" ring were replaced', the accumulator pressurized and satisfactorily leak checked, and returned to service.

Precautions to Provide for Reactor Safety During Repair: The work was performed per Procedure MCRD-3, Rev. 4.

All control rods were fully incerted.

h.

12/9/75 - Component: E-3 CRD accumulator had leakage from the locknut and spacer area between the gas and water bottles.

Cause of Malfunction: Normal wear of "O" rings and teflon rings.

Effect on Safe Operation: None, there was no withdrawn control rod incapable of scramming tr other incperable accumulator.

Corrective Action:

"O" rings and teflon rings were replaced.

A leak check after repairs was acceptable.

Precautions to Provide for Reactor Safety During Repair: Work was performed using MCRD-3, Rev. 5 3

CRD Pumps 7/31/75 - Component: No. 2 Control Rod Drive Pump.

a.

Cause of Malfunction: Excessive leakage from pump piston packing.

Effect on Safe Operation: The No.1 pump was in operation - no effect on plant operation.

Corrective Action: Added 2 ringa of packing in cylinder 2 and j

i 1 ring in cylinders 1 and 3 35

Precautions to Provide for Reactor Saic+.v_ During Repair: No. 1 i

pump in operation during maintenance.

b.

8/14/75 - Component: No.1 Control Rod Drive Pump.

Caur,e of Malfunction: Excessive leakage from piston packing gland.

Effect on Safe Operation: The No. 2 rod drive pump was in operation, so there was no effect on plant operations.

Corrective Action: Adjusted packing to stop leaking.

Precautions to Provide for Reactor Safety During Repair: The No. 2 pump was in operation.

10/9/75 - Component: The No. 1 CRD pump plungers needed repacking.

c.

Cause of Malfunction: Failure of pump plunger packing causing excessive leaking.

Effect on Safe Operation: None, one CRD pump may be removed from service during power operation.

Corrective Action: All three plungers were repacked and subsequently rechecked to establish proper leakage.

Precautions to Provide for Reactor Safety During Repair: Work was perfomed per Procedure MCRD-5, Rev.1.

The No. 2 pump was in service during repair.

d.

12/h/75-Component: No. 2 Control Rod Drive Pump packing leaked excessively.

Cause of Malfunction: Nomal wear.

Effect on Safe Operation: None, only one control rod drive pump is required during operation.

Corrective Action: Cleaned and repacked the No. 2 and 3 cylinders.

Precautions to Provide for Reactor Safety During Repair: Work was perfomed under MCRD-5, Rev.1 with the No.1 pump in service.

12/11/75-Component: Excessive leakage from No. 1 control rod e.

drive pump piston packing.

Cause of Malfunction: Nomal wear on packing.

Effect on Safe Operation: None - only one CRD pump is required for operation. Tightening packing does not remove the pump from service.

Corrective Action: Packing was tightened.

36

l l

/

Precautions to Provide for Reactor Safety During Repair: Not applicable.

f.

1/1/76 - Component: The No. 3 cylinder on the No. 2 control rod drive pump plunger.

Cause of Malfunction: Failure of pump plunger packing causing excessive leaking.

Effect on Safe Operation: None, one CRD pump may be removed from service during power operation.

Corrective Action: Installed new packing and checked for proper leakage.

Precautions to Provide for Reactor Safety During Repair: Work was performed using Procedure MCRD-5, Rev. 2.

The No. 1 pump was in service during repair.

4.

Control Rod Drive System Instrumentation 7/10/75 - Component: Accumulator Charging Header Pressure a.

Transmitter.

Cause of Malfunction: Broken internal connecting link.

Effect on Safe Operation: None. Redundant pressure indication i

is provided by the control rod drive pump discharge pressure transmitter.

Corrective Action: Replaced broken part and aligned indicator with system pressure.

Precautions to Provide for Reactor Safety During Repair: Not required. Transmitter provides redundant pressure indication only pressure control and annunciation is provided by separate instru-mentation.

(Mo 75-CRD-1013) 5 CRD Valves 8/30/75 - Component:

D-1 Selector valve a.

Cause of Malfunction: Leakage from valve.

Effect on Safe Operation: No effect, one control rod drive selector valve can be removed from service.

Corrective Action: Replaced seats, ball, stem and "O" rings in valve.

Precautions to Provide for Reactor Safety During Repair: Work

(

performed under Procedure MCRD-10 with all other rod drives in service.

37 g

y,

.v.-

w

b.

9/18/75 - Component:

C-2 Withdrav Rate Set Valve.

Cause of Malfunction: Leakage from the rate set valve.

Effect on Safe Operation: None, all control rod drives were scramable.

Corrective Action: Replaced rate set valve and subsequently leak tested the new valve.

Precautions to Provide for Reactor Safety During Repair: Work was performed per Procedure MCRD-9, Rev.1.

10/9/75 - Component: The B-5 Insert Selector valve was lessking.

c.

Cause of Malfunction:

"0" rings, seats and ball disc degradation causing air leakage.

Effect on Safe Operation: None, the scram function of the B-5 control rod was not impaired.

Corrective Action: The stem with "0" ring, backup ring, both seats and "O" rings, and the bass disc were replaced with new replacement parts. A subsequent leak test showed no leakage.

Precautions to Provide for Reactor Safety During Repair: Work was perfc.ned per Procedure MCRD-10, Rev.1.

The control rod was capable of scram during repair.

d.

12/7/75-Component: CRD D-2 Withdraw and Insert Rate Valves defective.

Cause of Malfunction:

F-1. F-2, F-3 filters clogged and two broken bolts on insert valve.

Effect on Safe Operation:

N' s, reactor in cold shutdown.

Corrective Action: Replaced D-2 withdraw and insert rate set valves, installed new bolts, and replaceE F-1, F-2, and F-3 filters.

Precautions to Pro ~ide for Reactor Safety During Repairs: Work was perfomed usin6 MCRD-9. Rev. 2 Reactor was in cold shutdown.

12/11/75 - Component: The B 6 Insert Selector Valve leaked.

e.

Cause of Malfunction: Nomal wear of ball valve parts.

Effect on Safe Operation: None.

Corrective Action: Replaced ball valve stem, stem backup ring, and "O" ring valve seats and valve "O" rings. A leak check after repairs was acceptable,

(

Precautions to Provide for Reactor Safety During Repai_r,: Repair was perfomed using MCRD-10, Rev.1.

38

6.

CRD Relief Valves 7/17/75-Component: No. 1 Pump Discharge Relief Valve.

a.

Cause of Malfunction: Leakage from valve.

Effect on Safe Operation: No effect, No. 2 pump was in operation.

Corrective Action: Replaced leaking valve with spare relief valve.

Precautions to Provide for Reactor Safety During Repag : Work perfomed under Procedure MCRD-llA, No. 2 pump in service during maintenance.

b.

8/21/75-Component: No. 2 Pump Relief Valve Cause of Malfunction: Excessive leakage from valve.

Effect on Safe Operation: None, one pump may be removed from service.

Corrective Action: Replaced the No. 2 pump relief valve and checked replacement for leakage. None was noted and the pump was returned for service.

Precautions to Provide for Reactor Safety During Repair:

Maintenance was performed under control of Procedure MCRD-llA with No.1 control rod drive pump in service.

(75 CRD-226 08) 9/4/75 - Component: Control Rod Drive Relief Valve on No. ' CRD pump.

c.

Cause of Malfunction: Leaking water through the valve seat.

Effect on Safe Operation: None.

Corrective Action: Removed valve from pump and installed new seat and disc. Reset the pressure settings and reinstalled on the pump.

Precautions to Provide for Reactor Safety During Repair: Work was controlled under Procedures MCRD 11-A, B, & C, Rev. O,1 & 1 respectively with one pump out of service.

d.

12/7/75 - Component: Leaking Flange / Nipple on No. 2 CRD Pump Relief Valve on December 6, 1975 Cause of Malfunction: Pipe nipple cracked in pipe threads due to fatigue caused by vibration palsations of the pump.

Effect on Safe Operation: None. One CRD pump can be removed from service.

Corrective Action: The relief valve flange / nipple was replaced.

j Precautions to Provide for Reactor Safety During Repair: Work 6

was performed utilizing Procedure 75 CRD 340-01, Rev. O.

39

7 M_iscellaneous 11/20/75-Component: Inability to withdraw the E 4 control a.

rod drive.

Cause of Malfunction: Not known at present.

Effect on Safe Operation: None, all control rods were fully inserted.

Corrective Action: In an attempt to gain withdrawal ability on E 4, the speed control manifold filters (F-1, F-2, and F-3) were replaced and the insert and withdrawal speed control valves (V-I and V-2) were replaced. After repairs to the speed control unifold it is still not possible to withdraw the drive.

i Precautions to Provide for Reactor Safety During Repair: Work was perfomed per Proceduro MCRD-9, Rev.1 with all control rods inserted.

I.

Reactor Vess'el No ma.ior maintenance was perfomed on this system during the report period.

J.

Post-Incident System No work on the post-incident system was perfomed during this report period.

K3 Fuel Handling Systems 10/2/75 - Component: Reactor Building Bridge Crane hydraulic a.

brake cylinder leaked fluid on the brake shoes.

Cause of Malfunction: Hydraulic brake cylinder fluid leakage.

i Effect on Safe Operation: None, the crane was not needed at the time but was available for use if needed.

Corrective Action: The hydraulic brake cylinder was replaced j

with a spare. It was checked for leaks, test operated, and adjusted.

Precautions to Provide for Reactor Safety During Repair:

Work was performed following the reactor building bridge crane service instructions for installation of brake cylinders.

b.

12/4/75-Component: The Reactor Building Crane main hoist cable was noted during Whiting Crane Corporation inspection on June 19, 1975, k

to be wearing.

40

Cause of Malfunction: Normal wear of cable.

(See also Wo 75-FHS-274-04 below)

Effect on Safe Operation: None, however the cable was not available for any capacity lifts at the direction of the Whiting Inspector.

Corre-tive Action: The new hoist cable was installed under the dirt..ons of the Whiting Inspector.

Precautions to Provide for Reactor Safety During Repair: Not applicable, the crane was not used for capacity lifts subsequent to 6/75 inspection.

12/h/75 - Component: Reactor Building Crane Sheave Guard rubbed c.

on the cable during crane operation.

Cause of Malfunction: The sheave guard could not be lubricated and consequently caused friction in the sheave which cocked the sheave c'ausing it to rub the cable.

Effect on Safe Operation: None, the crane operated adequately.

The rubbing, however, would cause wear to the hoist cable.

Corrective Action: The sheave guard was cut to allow adequate clearance of the hoist cable at the direction of the Whiting Inspector.

Precautions to Provide for Reactor Safety During Repair: Not applicable.

L.

Main Steam System 7/31/75 - Component: Turbine Bypass valve System Hydraulic Pump "A" a.

Cause of Malfunction: Pump would not pressurize high enough to clear low pressure alarm.

Effect on Safe Operation:

Generator trip could have resulted in reactor scram due to inability to dump steam to the condenser upon stop valve closing.

Corrective Action: Unloader was flushed clear and the setting was adjusted to 2,700 - 3,000 psig.

Precautions to Provide for Reactor Safety During Repair: Work performed at time of no threatening weather activities.

I 41 4

w

M.

Feed-Water System No maintenance performed on the safety-related portions of this system during the report period.

N.

Steam Drum 11/20/75-Component: Leakage from packing on upper east steam a.

drum instrument line valve.

Cause of Malfunction: Normal wear on valve packing.

Effect on Safe Operation: None, leakage was not excessive enough to effect operation. All rods were fully inserted at time of repair.

Corrective Action: Packing gland was taken up to the end of its travel on the valve indicated. No leakage was indicated.

In addition, packing on three other instrument line valves was taken up with ao leakage indicated after repair.

Precautions to Provide for Reactor Safety During Repair: Not applicable, the valve packing was taken up without removing the valves from service. All rods were fully inserted at time of repair.

O Station Power System 11/20/75 - Component: Several (9) station battery cells aboved a.

lower specific gravity readings than other cells.

Cause of Malfunction: The standard 24-hour overcharge was insufficient.

Effect on Safe Operation: None. Per the Exide Manual, electrolyte specific gravity reading limits are between 1.200 and 1.220 at 0

77 F.

Corrective Action: The station batteries were placed on over-charge for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Additional readings at that time showed all but one cell (51) had increased to the approximate value of other cells.

Precautions to Provide for Reactor Safety During Repair: Work was performed on batteries after overcharge using Procedure MSPS-5 (T30-20), Rev.1.

(

42 l

1

P.

Liquid Poison System 9/18/75 - Component: Temperature sensor on CV 4122 valve body a.

not indicating properly on temporary recorder.

Cause of Malfunction: Loose sensor cable connection at the temporary recorder.

Effect on Safe Operation: None. Temperature can be monitored with a portable potentiometer.

Corrective Action: Sensor cable properly secured to temporary recorder input terminals.

Precautions to Provide for Reactor Safety During Repair: CV 4122 valve body temperature was monitored with a portable potentiometer.

(MO75-LPS-1002)

Q.

Ventilating Air System 7/10/75 - Component: Containment Building Pipeway Dewpoint a.

Temperature Index.

Cause of Malfunction: Failed Deveell element.

Effect on Safe Operation: None. This unit is used for early detection of steam leaks in the containment pipeway. Backup measuresent is provided by the containnent outlet air dewpoint temperature sensor.

Corrective Action: Replaced the defective sensor, saturated and checked response. (MO 75-VAS-1012)

Precautions to Provide for Reactor Safety During Repair: Leak surveillance was maintained by the containment outlet air sensor.

b.

11/6/75-Component: The "A" heating and cooling unit for the Reactor Building has excessive leaks in the heat exchanger tube bundles.

Cause of Malfunction: Deterioration because of age.

Effect on Safe Operation: None, in that operation is not affected by the heat exchangers (it is an isolation component).

Corrective Action: Replaced the "A" heat exchanger with a new tube bundle.

Precautions to Provide for Reactor Safety During Repair: Work was perfonned under Procedure MVAS-1. Containment Integrity was protected.

I 43

c.

n/13/75 - Component: Reactor Building heating and co611ng "B" unit heat exchanger.

Cause of Malfunction: Age deterioration.

l Effect on Safe Operation: None in that the heat exchanger is an isolation component and operation is not affected by it.

Corrective Action: The tube bundle was replaced.

Precautions to Provide for Reactor Safety During Repair: Work was perfomed under Maintenance Procedure MVAS-1 with contaiment integrity being protected.

d.

11/13/75-Component: A check valve mounted in a line tapping off the piping downstream of CV-9460 had a minor steam leak which could potentially have lead to a degradation of contaiment integrity.

Cause of Malfunction: Nomal wear.

Effect on Safe Operation: None, the leak was minor and corrective action was complete prior to any degradation of contaiment integrity.

Corrective Action: The check valve was replaced.

Precautions to Provide for Reactor Safety Durind Repair: Work was perfomed at a time when no repairs were being perfomed on I

the heat exchangers within containment.

VI.

CHANGES, TESTS, EXPERIMENTS j

A.

Facility Changes Perfomed Persuant to 10 CFR 50 59 1.

Facility Change C-312 This change consisted of relocating the power source (120V a-c), to the Security Access Control System, from lighting panel 8L-5 breaker to Instrument and Control panel 3Y-6 breaker and increasing 3Y-6 breaker size from 15 amp 1;o 30 amp.

This change provides the Access Control System with an uninterrup-table power source.

The safety evaluation for this change concluded that the change would provide greater reliability of the Security Access Control System.

The 3Y power panel has sufficient capacity for additional load and

(

44

the wiring was properly sized for the increase in circuit capacity.

The Facility Change does not represent an unreviewed safety question as described in 10 CFR 50.59 2.

Facility Change C-314 This change involved replacing the 1/4" tubing in the Radiolytic Gas Sample System. That is downstream of NS-194 with 1/2" stainless steel tubing. The 3-way valve NS-194 was replaced with an elbow (valvewasnotused). Thereasonforthechangewasthatthe1/4" tubing was too restrictive for flow sampling.

The safety evaluation for this change concluded that the margin of safety would not be reduced. The change affects only the system's volume and the operating procedures will be unaltered. The safety factor is increased by eliminating a potential open path to the atmosphere.,This Facility Change, therefore, does not represent an unreviewed safety question as described in 10 CFR 50.59 VII. RADIOACTIVE EFFLUENT RELEASES A.

Introduction t

Releases of radioactive material, both to the atmosphere and to Lake Michigan frcxn January 1 to June 30, 1975, were well within the facility licensed limits and the NRC's regulations, particularly Title 10, Code of Federal Regulations, Part 20.

Table 1, Appendix A and Table 1, Appendix B, have been revised since the last Semiaviual Report to correct several typographical errors.

In addition, no'cle gas releases for January and June were incorrectly reported in the last semiannual. The reported releases were based on a 16-minute holdup time not the 30-minute holdup time that actually exists. The corrected values are contained in Appendix A of this re-port.

Table 2, Appendix B, of the 21st Semiannual Report, July 1,1974 to December 31, 1974, gave the " Percent of Applicable Limits" for the 12-month total liquid release as 22..OE+00%; this was a typographical error and should have been 2.10E+00%.

t 45

~B.

Gaseous Efnuent Releases to the atmosphere for the entire year totaled 5.06E+04 curies of fission and activation gases. This corresponds ti.; O.16% of the licensedlimitof1Ci/s. Particulate releases for the sert totaled 0.27 curie or O.13% of the licensed li-ait while halogen releases were measured to be 0.10 (t50%) curie or O.10 of the licensed limit. Gross alpha measurements on the particulate filter revealed that the year's release of alpha emitting nuclides totaled 2 35E-06 curies. Tritium releases for the year totaled 7.39 curies or 1.7E-05% of the limit based upon meteorological dispersion to the point of maximum ground concentration.

1.

Gaseous Effluent Calculational Methods A sample of off-gas is obtained weekly durim power operation and analyzed by gamma spectrometry for *six noble 6as radionuclides.

Based upon the mixture of the six nuclides, a stack release rate,

- which includes a total of 22 noble gas radionuclides, is detemined.

The stack release rate is based on a 30-minute holdup time for off-gas plus a 1% contribution from the turbine sealing steam system utilizing a 2-minute holdup. The 1% turbine seal contribution has the same dis-tribution of nuclides as the off-gas corrected for a 2-minute decay period. This is reflected in the monthly totals shown in Appendix A.

Activation gas releases are composed primarily of N-13 The rate of release is power-level dependent and is incorporated in the total monthly releases shown in Appendix A.

Particulate and halogen releases to the atmosphere are measuren by analyzing the particulate and charcoal filters weekly. These filters collect stack effluent continuously at a rate of three cubic feet per minute. Detemination of release rates in this manner assumes radio-

~

activity is continually being deposited unifomly throughout the week on the filters and, hence, a decay correction to the time of analysis

Xe-138, Kr-87, Kr-88, Kr-85m, Xe-135, Xe-133

\\

+,, _ - - - -

is applied, depending on the half-life of the nuclide observed. The g

net beta activity, as reported in Appendix A, represents the uniden-tified portion of the total activity present on the particulate filters (ie,grossbetaactivityminustheidentifiedisotopic activity). The net unidentified beta activity is corrected for decay based on a half-life of 27.7 years (ie, Sr-90).

Tritium releases to the atmosphere are calculated, based upon measure-ments made in the primary coolant and containment air and the follow-ing:

a.

Off-Gas - The average flow rate containing 90% radiolytic gas by volume at primary coolant tritium to hydrogen ratio and at 100% relative humidity is used to determine tritium releases both in vapor and molecular form.

b.

Turbine Sealing Steam - The design flo.i rate at 100% relative humidity and primary coolant tritium to hydrogen ratio.

c.

Containment Ventilation - The design flow rate and measured containment building tritium concentration.

The results of these celculations are also shown in Appendix A.

C.

Liquid Effluents Liquid waste releases totaled 1.41 curies of radioactive material for 1975 This release corresponds to 2.2% of Technical Specifications limits.

1.

Liquid Effluent Calculational Methods The release pathway to Lake Michigan for all liquid effluents is through the plant's condenser circulating water discharge canal. A flow of 48,000-52,000 gpn dilution for liquid effluents is obtained through the use of the condenser circulating water pumps, two at 24,500 gpm each and house service water pumps, two at 2,100 gpn each.

Each collected tank of liquid is sampled, analyzed for radioactive content, and discharged at a controlled rate to assure that permis-sible concentrations are not exceeded in the canal prior to dilution in Lake Michigan during the time of diucharge. Each sample is analyzed by geana spectrometry to identify as many of the ccxnponent nuclides as 47

1 i

possible (see f.ppenO x B for results).

Pemissible concentrations in the canal are determined frcut the following:

Ci MPCi ~ 1 g

where Ci is the concentration of the i_th isotope in the canal at the given concentration measured in the tank diluted by-the known canal a

flow rate. Those isotopes not identified by gamma spectrometry but measured by gross bete nalysis are presumed to be Sr-90 and released on that basis.

Tritium, Sr-89 and sr-90 analyses were perfomed by a contractor lab-oratory on monthly composite samples. Results of the June tritium analysis were not available for this report but will be included in the next Semiannual Report.

D.

Solid Wastes A total of approximately 157,000 curies of radioactive material was shipped off-site during the period covered by this report. Of the total, irradiated cobalt accounted for 154,000 curies, spent resin 1,006 curies, irradiated fuel rods 1,976 curies and miscellaneous solid radwaste 11 curies (see Appendix C).

E.

Environmental Dose Calculations Levels of radioactive materials in environmental media indicate that public intake is well below 1% of that which could result from con-tinuous exposure to the concentration values listed in Appendix B, Table II, 10 CFR, Part 20.

1.

Atmospheric Releases The integrated population dose, out to 50 miles, is shown on the following page. The following factors are utilized in the calcula-tion of the doses.

4 X/Q values for vhe five sectors are averaged over both stability a.

class and wind frequency.

b.

Doses are calculated for each of the 22 noble gas radienuclide.s and daughter products based on individual decay energies. Total dose is then the sununation of the individual nuclide contributions.

I i

j 48

_w,

c.

The 1975 population is estimated fr:n the 1970 Census of Populstion on a township basis corrected by the censue -determined State of Michigan growth rate of 1 3% per year and includes transient populationas1/4 residents. The total estimated 1975 population resides 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per day all year at the same location.

d.

The actual mixture foun?. during the weekly off-gas analysis is used for that week's releases and the total release is further corrected by daily gross measurements of off-gas.

e.

Site boundary doses are finite cloud shine doses. Semi-infinite cloud gemetry is utilized to calculate doses after the plume reaches ground level.

f.

No credit is taken for the meandering of the plume before it reaches the different annuli.

The maximum calculated radiation dose at the site boundary resulting from noble gas releases was 1 7 milli-Rems. The integrated dose to the population out to 50 miles was 1.4 man-Rems.

Doses from particulates and halogens releases as shown in Appendix A were negligible empared to that received fra noble gases due to the conservative limits in the plant Technical Specifications.

2.

Liquid Releases The nearest municipal drinking water supply intake is located in Charlevoix, Michigan, which is generauy upstream of the prevailing current flow in Lake Michigan at this location. However, since cur-rent patterns do occur that could, at times, carry the discharge water in the direction of Charlevoix, a population dose based upon this flow has been calculated. A conservative dilution factor of 800 is taken from the point of discharge to the City of Charlevoix based upon the report, " Big Rock Point Hydrological Survey, Great Lakes Research Division, University of Michigan, Special Repcrt No 9," by John C.

.tyers, 1961.

In addition, the poptiation dose is calculated to the entire po7ulation which receives its drinking water fra Lake Mich.igan, based on a unifom

(

49

i CAIn!IATED RADIATION DOSES FROM GASEOUS EFFIDENTS onnuary 1, 1975 to December 31, 1975 (pkn-Rem)

Sector Distance (Miles) 1 2

3 4

5 Total 1-2 Ibpulation 13 76 0

lo 0

99 Population Dose 4.73E-03 1.25E-02 3 09E-03 2.03E-02 2-3 Population 264 274 0

52 74 664 Population Dose 5 91E-02 3 20E-02 1.12E-02 1.61E-02 1.18E-01 3-4 Population 569 402 0

49 59 1,079 Population Dose 8.91E-oe 3.60E-02 8.13E-02 9 38E-03 1.43E-01 k-5 Population 731 3,387 0

104 0

4,222 Population Dose 3.87E-01 4.94E-02 1 34E-02 4.50E-01 5-10 Population 2,129 24 0

541 0

2,694 Population-Dose 1.05E-01 7.43E-04 3.25E-02 1 38E-01 10-20 Population 9,104 400 757 14,298 331.

24,89v Population Dose 1.1L5-01 3 10E-03 1.23E-02 2 30E-01 4.51E-03 3.64E-01 20-30 Pbpulation 9,776 3,550 1,927 4,683 331 20,267 Population Dose 3.65E-02 8.21E-03 9 88E-03 2 33E-02 1 38E-03 7 93E-02 30-40 Population 23,071 4,134 2,954 4,910 0

35,.69 Population Dose 3 72E-02 4.09E-03 6.44E-03 1.08E-02 5.85E-02 40-50 Population 41,320 9,004 5,949 12,258 0

68,531 Population Dose 3 50E-02 4.62E-03 6.73E-03 1.42E-02 6.0f>E-02 0-50 Population 86,977 21,251 11,587 36,905 795 157,515 Population Dose 8.68E-01 1 51E-01 3 54E-02 3.47E-01 3 14E-02 1.43E+00 Site Boundary Dose 1.40 0 94 1.66 1.60 (Milli-Rem)

I 50 l

concentration, resulting from plant releases, throughout Lake Michigan. Also, radiation dose to human populations can ex: cur as a result of plant releases through the consumption of 1-ish caught in Lake Michigan.

Utilizing the measured values of radionuclides released as shown in Appendix B, the following fomula, and the standard man model, drinking water doses can be calculated as follows:

I f

C D, =

r (LimitingDoseRen/Yr)

{

Ll where:

D, is the individual dose in Rec /yr, C is the average concentration in Lake Michigan of the 1

individualnuclidesmeasured,inpCi/ml, MPC is the concentration of each nuclide measured, required to

'produce the li:niting dose at continuous intake in pCi/ml and the limiting dose is ue dose produced at continuous exposure to MPC concentrations.

In calculaM.ng ingestion dose frcan the consumption of fish, an equa-tion similar to the one used for drinking water dose is used except that a. standard daily diet of 50 grams of fish flesh is used in con-trast to the 2,200 ml of fluid consumed daily by the standard man.

This, in effect, increases the MPCi by 2,200/50 or M.

The calculation of individual doses, both from drinking water and con-suming fish, are per the previous fomula while integrated population doses in man-Rem are calculated utilizing the following parameters:

a.

For drinking wr.ter, the individual doses ere sununed over the entire penult. tion that receives its drinking water from Lake Michigan wit. discharge canal flow appropriately mixed with the lake.

Th'.s is approximately 10 million people of which approximately 7 million reside in the Chicago metropolitan area.

b.

The population dose due to drinking water to Charlevoix residents is based on a population of 3,500 people.

(

51

i c.

For fish consumption, the average concentration in Lake Michigan water, resulting from plant releases, is used with a bioaccumula-tion factor to detemine the average concentration in fish, d.

Fish do not reside continuously in the discharge canal but migrate Population doses based upon drinking water from the Charlevoix munici-pal system were 0.008 man-Rem and total Lake Michigan drinking water consumption population dose was 0.26 man-Rems. The consumption of alJ of the Lake Michigan fish harvested resulted in a population dose of 0.11 man-Rem.

4 As a measure of total environmental impact, the radioactive liquid releases from the plan. are averaged over the entire lake and then used to determine the 1 pulation dose fran fish caught throughout the entire lake and tott t water consumed from the lake.

Both of the dose c alculations are conservative in that:

a.

Equilibrium is not obtained in the human body for most isotopes released.

b.

No credit is taken for pree'.pitation and deposit in sediment or uptake by life foms other than fish, c.

No credit is taken for radicactive decay which for I-131 is significant.

Results are shown in the following tables.

ERG Special Report No 2, " Trace Element Distributions in Lake Michigan Fish: A Baseline Study With Calculations of Concentration Factors and Equilibrium Radioisotope Distributions," March 1973 k

52

CONSUMD JWER COMPANY Big Rock Nuclcar ibwar PITLnt Calculated Ev11ation Doses Frcza Liquid Effluents - Population Drinking Water Dose From1/1/75to12/31/75 Avg Concentration (Ci/MICi)

Population Populatior. Dose Critical Curies in Lake Michigan MPDI Dose Charlevoix$

MI Vector Isotope MPC Organ Released (uci/al)

Ci/MPCi (mrem /Yr)

(Man-Rem)

(Man-Ren Water Zn-65 1.OOE-Oh Whole Body 0.0274 4.71E-15 4.71E-ll 2.86E-08 0.00029 8.92E-06 I-131 3.OOE-07 Thyroid 0.0077 1.60E-15 5.33E-09 2.66E-06 0.02665 8.32E-04 Cs-134 9.OOE-06 Whole Body 0.1266 2.64E-14 0.93E-09 1.47E-06 0.01465 4.58E-04 Cs-137 2.OOE-05 Whole Body O.4072 8.48E-14 4.24E-09 2.12E-06 0.02121 6.62E-04 BaLa-140 2.OOE-05 GI Tract 0.0030 6.22E-16 3.11E-11 4.66E-08

0. COO 47 1.46E-05 Co-58 1.OOE-06 GI Tract 0.0002 3 99E-17 3 99E-11 5 99E-08 0.00060

- 1.87E-05 Co-60 3.OOE-05 GI Tract 0.1394 2 90E-14 9.68E-lO 1.45E-06 0.01452 4.53E-04 Fe-59 5.OOE-05 GI Tract 0.0011 2.29E-16 4.58E-12 6.87E-09 6.87E-05 2.15E-06 Mn-54 1.OOE-04 GI Tract 0.0417 8.69E-15 8.69Z-11 1 30E-07 1.30E-03 4.07E-05 sr-89 3.OOE-06 Bone 0.0029 6.04E-16

2. ole-lO

1. ole-07

1. ole-03 3.15E-05 g

Sr-90 3.OOE-07 Bone 0.0008 1.67E-16 5.56E-10 2.78E-07 2.78E-03 8.67E-05 Others 3.OOE-06 Whole Body 0.6520 1.36E-13 4.53E-08 2.26E-05 2.26E-01 6.95E-03 Total Whole Body 2.62E-Ol 8.08E-03 Thyroid 2.67E-02 8.32E-04 GI Tract 1.70E-02 5.29E-04 Bone 3.79E-03 1.18E-04

CONSUMERS POWER COMPANY Big Rock Nuclear Power Plant Calculated Radiation Doses From Liquid Effluents - Fish Consumption Dose From 1/1/75 to 11/30/75 Avg Concentration Avg Concentration Critical Bioaccumulation in IAke Michigan in Fish (CFI/MPCI)MFDI Population Dose Vector Isotope MPCi Organ Factor (u C1/ml)

(.Ci/g)

(mrem /W)

(Man-Rem)

Fish zn-65 4.40E-03 Whole Body 900 5.71E-15 5.14E-12 5.84E-07 0.00035 I-131 1 32E-05 Ihyroid 500 1.60E-15 7 99E-13 3 03E-05 0.01799 Cs-134 3 96E-04 Whole Body 2360 2.64E-14 6.22E-ll 7.86E-05 0.04668 Cs-137 8.80E-04 Whole Body 2360 8.h8E-14 2.00E-10 1.14E-04 0.06757 BaIa-140 8.80E-04 CI Tract 365 6.22E-16 2.27E-13 3.87E-07 0.00023 Co-58 4.40E-05 GI Tract 330 3.99E-17 1 32E-14 4.49E-07 0.00027 Co-60 1 32E-03 GI Tract 330 2.90E-14 9 58E-12 1.00E-05 0.00647 Fe-59 2.2E-03 GI Tract 260 2.26E-16

!.88E-14

4. ole-08 2.40E-05 Mn-54 4.4E-03 GI Tract 280 8.69r-15 2.43E-12 8.30E-07 4.97E-04 Sr-89 1.30E-04 Bone 18 6.04E-16 1.09E-14 4.18E-08 2 51E-05 sr-90 1 3E-05 Bone 18 1.67E-16 3.01E-15 1.16E-07 6.93E-05 4'-

Others 1 3E-04 Whole Body 80 1.36E-13 1.09E-ll 4.18E-05 2.51E-02 Tocal Whole Body 1.40E-01 Thyroid -

1.80E-02 GI Tract 7.49E-03 Bone 9.44E-05 l

Note: MPCi = MPC (Water)

2200/50 Population dose based on annual catch of 2.3E+7 pounds consumed at 50 Gm/ day / person.

1

VIII. DWIRONMENTAL MONITORING 4

A.

Environmental Survey i

Environmental levels of radioactivity as found in the v of the plant were composed almost entirely of naturally occurr g radioactive i

materials. In the vicinity of the circulating water discharge canal, radioactive material of plant origin has been found. These materials occurred primarily in aquatic organisms. The levels of radioactive materials, however, were extremely low and are of no significance to the health and safety of the organisms or the public. Further, the levele of radioactive material found in the resident biological com-munity are consistent with levels found in previous years and show no upward trend.

The environmental surveillance program includes continuous campling of air for particulate and halogen activity at seven locations to deter-mine increased concentrations, if any, of radioactivity of plant origin.

Included are two background sample locations at Traverse City and Boyne City, Michigan, about 50 miles south-southwest and 20 miles southeast of the plant, respectively.

Thermoluminescent dosimeters (TLD), placed at each of these locations plus six additional locations on the site boundary, measure direct dose in the environment. In addition to the exposure received in the field, dosimeters also receive an exposure in transit to and from the con-tractor laboratory. The total monthly doses which include the "in transit" dose, measured by the dosimeters have been evaluated to deter-mine if any dosimeter or set of dosimeters is receiving doses signifi-cantly different from the rest. Part of the evaluation included the plotting of the dosimeter readings versus cumulative percent of the ranked data on log-nonnal paper. If the plotted data results in a straight line, the data then fits a log-Gaussion distribution. Data points not falling on this line do not belong to the same distribution as the other dosimeters. Or stated simply, these t'osimeters have been exposed to either a lesser or an addit'.onal amount of radiation.

f 55

The examination of the TLD results shows that the monthly readings, after separating the two different dosimeter series utilized, all fit a log-Gaussion distribution. Therefore, the site boundary dose, as measured by TLD, is not significantly different from the dose measured by the off-site TLD. The monthly averages and standard deviations (' sigma) for those distributions are given in Appendix D.

Air samples gathered enn+1nuously and analyzed weekly at the stations shown in Appendix D showed no difference, at the 95% confidence level, in the level of radioactivity measured at those s+1tions close to the site and those remote from the site. Both particulate filters and car-bon cartridges are used to measure potential concentration of radio-active materials resulting from plant opere.tions. From average meteorological dispersion conditions, the following maximum concen-trations can be calculatt2:

Particulates=(1.2pCi/s)x(5.0E-14s/cc)(990E-04)

Halogens =(1.2uCi/s)x(5.0E-14s/cc)(1.32E-03)

These compare to the minimum detectable activity values and normal background concentrations as follows:

Maximum Calculated Normal Concentration Mini== Detectable

Background

Release uCi/cm3 Activity uCi/cm3 Activity uCi/cm3 Particulates 5 9E-17 1.0E-14 7.0E-14 Halogens 7 9E-17 2.0E-14 Hence, the negative data obtained in the program was expected.

During the year, only four air samples showed detectable concentrations of I-131. The site air samples for the collection periods of April 17 through 24 and May 1 through 8 bad I-131 concentrations of 0.02 t 0.01 l

3 pCi/m, respectively. The Nine Mile Point air samples and 0.03

0.01 for October 16 through 23 and October 23 through 30 both had I-131 con-3 centrationsof0.03t0.01pCi/m. All four samples had low airflow volumes and, therefore, the values probably represent statittical vari-ations not real ambient concentrations.

t l

56 1

At the Big Rock Point Plant, daily composite condenser circulating water inlet and canal water discharge samples are taken and analyzed for radioactive content. In addition, a monthly composite of these samples is analyzed for radioactive content (gross beta and tritium).

These results plus site well water and Charlevoix drinking water are shown in Appendix D.

Additional aquatic samples are taken and analyzed during the spring and fall. These results are also tabulated in Appendix D.

The release of 1.4 curies of liquid radioactive effluents (less tritium and noble gases) results in an average concentration in the dischr.rga canal of 1.4E e pu1/ml. The analysis of discharge canal water should, therefe -

'.ndicate an increase of radioactive material in discharge canal water samples since the mini-m detectable activity forgrossbe.tameasurementsisabout2.0E-09uci/mlorabmtseven times lower than the predicted discharge concentration. The results shown plotted in Appendix D indicate an average measured discharge concentrationofabout6.OE-09pci/mlaboveintakeconcentrationor roughly two times lower than the predicted concentration based on effluent analyses.

Also contained in Appendix D are a sinmag of the sample analysis schedule and the high, low and average concentraticn for the highest average sampling location.

IX.

OCCUPATIONAL PERSONNEL RADIATION EXPOSURE

These followin6 exposures were tabulated from pocket dosimeter records, radiition protection logbouks. veekly radiation exposure record work sheets and high radiation area work sheets. The tabulation is considered to be 10% high since the pocket dosimeters normally read 10% higher than the film dosimeters.

The report period is July 1, 1975 through December 31, 1975 OPocket dosimeter accumulations unless otherwise noted.

57

Total Exposure Operations mR Footnote Routine Plant Surveillance 8,116 (1)

Radwaste or Fuel Pool Filter Change 276 (5)

Pipe Tunnel Inspection 249 (1)

-Decontaminated Fuel Pool Heat Exchanger 28 (h)

Steam Drum Area Steam Leak Repair 755 (h)

Clean Up Demin Pit Inspection 135 (1)

Recirc Pump Room 957 (1)

Opened Steam Drum Door to Listen For Leaks 12 (1)

Ex Level Inspect Fire Hose 6

(1)

Turbine Steam Seal Regulator Repair h2 (h)

Rx Level Treat II Cask 35 (2)

To Verify That Locked Valves Were Locked in Recire Pump Rm & Steam Drum Area 56 (1)

Shutdown Em

~

166 (1)

Resin Moving Radwaste 39 (5)

Recire Pump Repair Valve 35 (4)

Valving in Shutdown Rm 305 (2)

Total 11,212 Shift Supervisors Routine Plant Surveillance 1,739 (1)

Main Condenser Area - Inspection 573 (1)

Recire Pump Em. - Inspection 1,915 (1)

Turbine IPR - Surveillance 22 (1)

Valve Leak Inspection - Drum Area 750 (1)

High Pressure End Turbine Repair.

35 (1)

Total 5,03h All Remaining Supervisory Personnel Routine Plant Surveillance 1,834 (1)

Main Condenser Area - Inspection 332 (1)

High Pressure End Turbine - Inspection 108 (1)

Turbine IPR - Inspection 132 (1)

Clean Up Heat Exchange Room - Inspection 138 (1) t Total 2,544 58

Engineering Personnel i

Routine Plant Surveillance 3.325 (1)

Clean Up Pit Entry 286 (1)

High Pressure End Turbine _ - Inspection 24

.(1)

Main Condenser Area - Inspection 51 (1)

Turbine IPR - Inspection 116 (1)

Rx Deck - Inventory Fuel Rods 100 (6)

Recire Pump Em. - Inspection 16 (1)

Total 3,918 Total Exposure Maintenance Personnel mR Footnote Routine Plant Surveillance 5.467 (1)

Cleaned Floor in Room 444 40 (2)

Disassembling Recire Pump Seal in 121 63 (2) a Rx Level Work Uith Vacuum Hose 100 (4)

West Side Of Turbine Removing Security Gratin'g 17 (4)

Stud Tensioner Repair 30 (2)

Fuel Pool Hx Cleaning (Decontamination) 767 (4) l Repair of Feed H O Heater Drain Valve 44 (4) 2 Radwaste Tank Rm Floor Valve Repairing 75 (4)

Repair of Accumulator Seals 30 (4)

Sclid Radvaste Disposal 757 (5)

Change Rod Drive Filter 20 (2)

Resin Sluice Valve Pit 20 (4)

Turbine Valve Repair - Steam Regulator 50 (4)

Stock East Air Flow Test 10 (4)

Shutdown Hx Em - Inspection 12 (1)

Shipping Drums to Nuclear En61neering Co.

73 (5)

Turbine IPR Repair 391 (4)

Treat II Cask Work 60 (2) d Repair of Radwaste Strainer 284 (4)

Sipper Can Rack Removal 387 (4)

Radwaste Pump Strainers 50 (4)

Radwaste Pump Suction Line 51 (4) l 59

-F 4-y-

,y,,-

Total Exposure Maintenance Personnel mR Footnote Steam Drum Area - To Tighten Packing on Valves 50 (h)

Reactor Deck Equipment Check 50 (2)

Reactor Working Platfom Repair 30 (h)

Clean Up Pump Em Exchanger Wiring 255 (h)

Pipe Tunnel Steam Leak Repair 127 (4)

Reactor Deck - Welding on Working Platform 40 (4)

Chemical Waste Receiver Tank Repair 6,533 (4)

NDP Dye - Pen on Working Plattom 94 (h)

Total 15,977 Total Exposure Health Physics and Chemistry mR Footnote Routine Plant Surveillance 3,663 (1)

Main Condenser Area - Inspection 547 (1)

Cleaned Floor in Rm hhh - Monitoring 20 (7)

Cleaninr Fuel Pool Hx - Monitoring 99 (7)

Steamicek Feed H O Heater Drain - Monitoring 58 (7) 2 Steam 1 rum Leak in Valve - Monitoring 169 (7)

RW Tank Em Floor Valve - Monitoring 20 (7)

RW Cask to Solid RW - Monitoring 45 (7)

Calibration of Area Monitors 83 (2)

Recire Pump Rm - Check fcr Leak RCW System 490 (1)

Clean Up Demin 50 (7)

Fuel Fool Cask to Radweste 25 (7)

Steam Drum Door to Listen for Leaks 10 (1)

Recire Pump Em - Inspection 480 (1)

Steam Drum Steam Leak 580 (7)

High Pressure End Turbine Repair 20 (7)

Shipping Drums to NECo.

22 (7)

Treat II Cask 20 (7)

Turbine IPR 118 (7)

Canning 4 Fuel Rods 35 (7)

Pipe Tunnel Ht. Drain Repair 36 (7)

Transfer of Solid Radwaste 50 (7) i I

60

Total Exposure i

Health Physics and Chemistry mR Footnote Sipper Can Back Removal 130 (7)

Steam Drum Area Tighten Packing on Valves 20 (7)

Grand Tour (Leak Inspection) 60 (1)

Rod Drive Rm RD Position Indication

'28 (7)

Recire Pump Rm - Inspection 50 (1)

Pipe Tunnel Repair Steam Leak 32 (7)

Steam Drum Repair Valve hk (7)

Clean Up Pump Uiring 65 (7)

Radvaste Tank Rm - Inspection 20 (1)

Chem Tank Repair 3,067 (7)

Recire Pump Em - Inspection 70 (1)

Total 10,226 Total Exposure Instrument and Control mR Footnote Routine Plant Surveillance 1,326 (1)

Fuel Pit Hx Rm and Pumps hk (h)

Turbine IPR 133 (4)

Pipe Tunnel Dev Cell Repair 42 (h)

Rod Drive Rm - RD Position Indication 50 (h)

Clean Up Hx Rm Wiring 1,125 (4)

Total 2,720 QA and QC Personnel btinePlantSurveillance 426 (1)

All Other Personnel General Electric Personnel on Turbine IFR 382 (b)

Neukirk on Security System 334 (2)

Catalytic on Rapid Depressurization System 7,383 (4) 1 61

r Footnotes (1) Reactor Operations and Surveille.nce This includes all routine surveillance inspection, (except for inservice inspections).

(2) Routine Maintenance All maintenance that is scheduled. This does not mean however, that particular date(s) are established to qualify for being scheduled.

This meintenance would have been plamed on and would occur at least yearly.

(3) Inservice Inspection Inspections nomally performed by QA, NIYr personnel, and outside contracted personnel. Nomally inspections conducted during an outage of systems and piping that cannot be checked while at power.

NOTE: Inservice Inspection, Technical Definition.

An inspection perfomed in accordance with the rules of the ASME B&PV Code Section XI on nuclear power plant components.

(4) Special Maintenance All maintenance that has not been scheduled for in any way. This maintenance would not have been planned for in advance and nomally could not have been predicted.

(5) Waste Processing Includes any work with solid or liquid radwaste. Movement of casks and liners. Radvaste or fuel pool filter changes. Resin moving. Bailing of low level radvaste.

(6) Refueling All work with~ fuel or reactor components perfomed by the Operations Department in the reactor and pool area.

(7) Monitoring This would include all monitoring done by Health Physics personnel.

(

62

OCCUPATIONAL EXPOSURE Number of Persons Within Exposure Range mrem **

Dose 7/1/75-7/31/75 8/1/75-8/31/75 9/1/75-9/30/75 0-100

Maint 11 Oper IT Maint 12 oper 19 Maint 10 Oper 15

.Supy 22 Tech 10 Supv 20 Tech 10 Supv 22 Tech 10 Others 25 Others 25 Others 46 101-500

Maint 4 Oper 3

Maint 3 Oper 1 Maint 5 Opec 5

Supy 0 Tech 1

Supy 3 Tech 3 Supv 2 Tech 2

Others 1 Others 0 Others 0 501-1250

Maint 0 Oper 0

Maint 0 Oper 0 Maint 0 Oper 0

.Supy 0 Tech 0

Supy 0 Tech 0 Supy 0 Tech 0

Others 0 Others 0 others 0 1251-2500 Maint 0 Oper 0

Maint 0 Oper 0 Maint 0 Oper 0

Supv 0 Tech 0

Supv 0 Tech 0 Supy 0 Tech 0

Others 0 Others 0 Others O

> 2500 Maint 0 Oper o

Maint 0 Oper 0 Maint 0 Oper 0 Supv,

O Tech 0

Supy O.

Tech 0 Supy 0 Tech 0 Others 0 Others 0 Others 0 Total Number of People Badged 94 97 117 mrem Dose 10/1/75 - 10/31/75 11/1/75-11/30/75 12/1/75-12/31/75 0-100

Maint 12 Oper 14 Maint 10 Oper 18 Maint 4 Oper 13 Supy 21 Tech 5

Supv 23 Tech 10 Supy 24 Tech 4

Others 42 Others 73 others 101 101-500 Maint 6 Oper 6

Maint 6 Oper 1

Maint 2 Oper 8

Supy 3 Tech 5

Supv 0 Tech 1

Supv 2 Tech 3

Others 0 others 7 Others 12 501-1250 Maint 1 Oper 0

Maint 0 Oper 0

Maint 5 Oper 0

Supy 0 Tech 1

Supy 0 Tech 0

Supy 0 Tech 3

Others 0 Others 0 Others 1 1251-2500 Maint 0 Oper 0

Maint 0 Oper O

Maint 1 Oper 0

Supy 0 Tech 0

Others O Others 0 Others 0 7 2500 Maint 0 Oper 0

Maint 0 Oper 0

Supy 0 Tech 0

Others 0 Others 0 Others 0 Total Number of People Badged 128 142 178 Others include office secretaries, rieneral Office personnel, contract personnel, vendors, plant guards, Region repairmen (other than from Traverse City) visitors

[

and Quality Assurance Dept.

Maint includee Region repairmen from Traverse City.

- Film dosimeters 63

NUMBER OF PERSONNEL AND MA.'

24 BY WORK AND JOB FUNCTION AT THE BIG ROCK POINT NUCLEAR GENERATING PLANT l

T/1/75 THROUGH 12/31/75 NumEe'r of Personnel ( > 100 mrem)

Total mrem Station Utility Contract Work Station Utility Contract Work and Job Function Employees Employees

. Workers &

Employees Employees Workers &

Othere Others REACTOR OPERATIONS & SURVEILLANCE

. Maintenance Personnel 13*

5,571 Operating Personnel 20*

9,697 Htalth I5ve'.es Personnel 6*

5 390 Supervisory Personnel 7*

7,578 Engineeriog sersonne1 4*

3,818 Instrument & Control Personnel 5+

1,326 Clerical Personnel 0

0 QA & QC 3

426 NI/f Personnel 0

0 9

All Other Personnel O

O ROUTINE MAINTENANCE M;intenance Personnel 6*

263 Operating Personnel h*

440 Health Physics Personnel 2*

83 Supervisory Personnel Engineering Personnel Instrument & Control Personnel Clerical Personnel QA & QC Personnel NUP Personnel All Other Personnel CEstimated

NUMBER OF PERSONNEL AND MA.~

Ji BY WORK AND JOB FUNCTION AT THE BIG ROCK PODIT NUCLEAR GENERATDIG PLANT 7/1/75THROUGH 12/31/75 Humber of Personnel ( > 100 mRcm)

Total mrem Station Utility Contract Work Station Utility Contract Work and Job Function Employees Employees Workers &

Employees Employees Workers &

Others Others INSERVICE HISPECTION NONE Maintenance Personnel Operating Personnel H:alth Physics Personnel Supervisory Personnel Engineering Personnel Instrument & Control Personnel Clerical Personnel QA & QC Personnel N M Personnel g

All Other Personnel SPECIAL MAINTENANCE M2intenance Personnel lo*

To*

9,311 Operating Personnel 6*

860 Health Physics Personnel Supervisory Personnel 7*

Engineering Personnel 7*

Instrument & Control Personnel 5*

1,394 Clerical Personnel 3*

QA & QC Personnel 94 N M Personnel

,Mai t., Supv.

, 82 GE on IPR 3

All Other Personnel Eng., Clericals 334 Neukirt QA) i,7303 Cui.ul tic

/

oEstimated

NUMBER OF PERSONNFL AND MAN-nEM BY WORK AND JOB FUNCTION AT THE BIG ROCK POINT NUCLEAR GENERATING PLANT T/1/75THROUGH 12/31/75 Number of Personnel (> 100 mrem)

Total mrem Station Utility Contract Work Station Utility Contract Work and Job Function Employees Employees Workers &

Employees Employees Workers &

Others Others WASTE PROCESSING Maintenance Personnel 3*

830 Operating Personnel 1*

315 HIslth Physics Personnel Sdpervisory Personnel Engineering Personnel Instrument & Control Personnel Clerical Personnel QA & CE: Personnel m,

lNITfPersonnel i All Other Personnel REFUELING Mnintenance Personnel Operating Personnel

' H alth Physics Personnel Supervisory Personnel Engineering Personnel 1

100 (Inventory of Fuel Rods)

Instrument & Control Personnel Clerical Personnel QA & QC Personnel NI7f Personnel All Other Personnel CEstimated

NGGER OF PERSONNEL AND MAi E BY WORK AND JOB FUN.XION AT 'DE BIG ROCK POINT NUCLEAR GENERATING PLA?T 7/1/75 THROUGI 12/ '1/75 Number of Personnel () 100 mrem)

Total mrem Station Utility Contract Work Station iUtility Centract Work and Job Function Employeen Employees Workers &

Employees Employees Workers &

Others Others MONITORING Maintenance Personnel Operating Personnel 4,753 Health Physics Personnel 6

Supervisory Personnel Engineering Personnel Instrument & Control Personnel Clerical Personnel QA & QC Personnel NITf Personnel All Other Personnel Tol'AL EXPOSURE Maintenance Personnel 13 70*.

15,883 (Maint.,

{8,099 "Eg Operating Personnel 20 11,212 Health Physics Personnel 6

10,226 Supervisory Personnel 14 7*

7,578 Engineering Personnel 11 7*

3,918 Instrument & Control Personnel 5

2,720 Clerical Personnel 3*

QA & QC Personnel 3

426 NITf Personnel 94 All Other Personnel 72 52,268 94 8,099 OEstimated

i X.

RADIOACTIVE LEVEIS IN PRINCIPAL FLUID SYSTEMS Minimum Average Maximum A.

Primary Coolant pCi/mpayaterFiltrate Reactor

-2 1.6 x 10' 5 1 x 10' 7 3 x 10 Reactor Water Crud pCi/ml/ Turbidity Unit (

2.4 x 10' 6.4 x 10' 5 8 x 10-Iodine Activity, pCi/ml(b) 5 8 x 10 1.8 x 10 1,5 x yg

-3

-2

-1 B.

Reactor Cooling Water System ReactorjoolingWater

-3 2.6 x 10 8 3 x 10

-2

,uCi/mp a 7 3 x 10 C.

Spent Fuel Pool FuelStoragePool("}

5 1 x lo-5 0 x lo 3 5 x 10

-I

-5 1.0 x 10-9 2 3 x 10-6 3 5 x 10-5 Fuel Storage Pool Iodine (a)A counter efficiency based on a decay scheme consisting of one gamma photon per disintegration at 0.662 MeV used to convert count rate to microcuries.

All count rates were taken two hours after sampling.

(b) Based on efficiency of Iodine-131two hours after sampling.

(

Based on APHA turbidity units and 500 ml of filter sample.

l t

i 68

Revised 2-15-76 APPENDIX A. TABLE 1 REVISYD Consumers Power Compar.y Big Rock Point Plant, D:eket 50-155 Atmospherie Release of Radioactive Material Six-Month Units January February March April May June Total Total Noble Gases Curies 1.26F+03 3.6TE+03 k.93E+03 Total Halogens 1 98E-02 3.36E-Oh 1.2kE-Ok k.23E-04 2.0TE-02 Total Particulates (8, y) 9.63E-03 k.68E-Oh 2.lTE-Oh 1.kSE-03 9.k1E-Ok 6.89E-Ok 1.3kE-02 Total Tritium 5.62E-01 1.38E-01 1 53E-01 1.kAE-01 1 53E-01 T.23E-01 1.88E+00 Total Particulates - Oross Alpha 2.00E-07 1.0TE-07 2.kBE-07 1.TSE-07 1.78E-04 1.69E-07 1.08E-06 Maximum Noble Gas Release Rate aci/s 1.kOE+03 2.58E+03 2 58E+03 Percent of Technical Specifications Noble Gases k.70E-02 1.k2E-01 T.03E-03 Halogens 2.06E-01 1.16E-02 3.8kE-03 k.66E-03 3.86 4C2 Particulates (8, y) 8.56E-02 1.16E-02 1.5TE-02 2.83E-02 2.13E-02

2. bE-02 3 20E-02 IIotopes Released Curies Halogens I-131 2.25E-03 3 35E-ok 1.2kE-Oh 5.22E-05 2.76E-03 I-133 1.75E-02

3. TIE-Oh 1.79E-02 Particulates Cs-13h 2 57E-05 5.01E-06 1.06E-Oh 3.58E-05 1.6TE-Oh Cs-13T 5 19E-Oh

2. 36E-04 5.61E-05 1.23E-05 1.65E-Ok k.6TE-05 1.0LE-03 Bata-lho 6.99E-03 k.1TE-06 3.13E-06 7.00E-03 Mn-5h 2.62E-Oh 1.88E-06 3 90E-05 3.kSE-05 1.1TE-05 3.k9E-OL Co-60 2 97E-Oh 1.6TE-Oh 1.18E-Oh 1.22E-03 k.6TE -Oh 6.2kE-05 2 33E-03 Co-58 5.28E-05 5.28E-05 Net Unidentified Beta 1.k8E-03 5.k5E-05 3 93E-05 1.75E-Ok 1.68E-Oh 5 33E-Oh 2.k5E-03 Noble Cases Xe-138 2.02E+02 T.C8E+02 9 10E+02 Kr-87 1.70E+02 T.33E+02 9 03E+02 Kr-88 1.k6E+02 k.00E+02 5.k6E+02 Kr-85m 6.81E+01 2.02E+02 2.70E+02 Xe-135 2.9hE+02 8.2TE+02 1.12E+03 Ie-133 1.10E+02 2.63E+02 3 73E+02 Xe-lk3 Kr-94 Kr-93 Xe-1k1 Kr-92 Kr-91

<1

<1 Xe-1k0

<1 2.39E+00 2.93E+00 Kr-90 8.kKE+00 2.$1E+01 3.36E+01 Xe-139 1.25E+01 3.72E+01 k.97E+01 Kr-89 1.83E+01 5.hkE+01 7 2TE+01 Xe-13T 3.19E+01 9.k6E+01 1.2TE+02 Xe-135m 6.63E+01 1 97E+02 2.63E+02 i

Kr-83m 3.85E+01 1.1kE+02 1.53E+02 Xe-133m 3 37E+00 9 99E+00 1.3kE+01 Xe-131m

<1

<1 Kr-85

<1

<1 N-13 8.38E+01 1.09E+02 1 93E+02

(

69 P00R ORIGINAL

Pevised 2-15-76 APPEC TI A. TAB *E 2 Constaners Power Company Big Ro:k Point flant. Docket 50-155 Atmospheric Release of Radioactive Material Six-% nth Yearly Unito July Ag2st Bertember October hvember Decen.her Tot L1 Total Total Noble Casco Curies 6.38E+03 6.1TE+03 5.k9E+03 5 98E+03 1.20E+0h 9.65E+03 b.57Dok 5 06E+0h Total Raiogens b.20E-Ok 8.36E-Oh 3.63E-Ok 3.22E-03 6.86E-03 2.35E-01 2.LTE-01 2.67E-01 Tott.1 Particulates (6. y) 2.10E-03 2.69b03 1.09E-03 8 32E-C3 6.69E-02 3 5kE-03 6.k6 bO2 9.BOE-02 Tots 1 Trititas 9.k3E-01 9.k3E-01 9 11E-C1 9.kkE-01 8.72E-01 8.96E-01 5 51E+00 T.39E+00 Total Particulates - Cross Alpha 1.28E-07 2.06E-07 1.13E-07 3 78E-O

9 11E-08 3.01E-07 1.28E-06 2 35E-06 Maziams Noble Gas Release Rate aci/s 3.k6E+03 2.63D03 2.kBE+03 5 2kE+03 1.10E+0k T.72E+03 1.10E+0L 1.10E+0k Prresnt of Technical Specifications 1

Noble Gases 2 38b O1 2.30E-01 2.12E-01 2.23bO1 k.6LE-01 3.60E-01 2.89E-01 1.61E-01 Halogens 5.28E-03 1.0kbO2 5 61E-03 b.12E-02 1.88E-01 1.09E+00 2.23501 1 32E-C1 Ptrticulates (8. 3) 3 21 bO2 b.01E-02 3.76E-02 2.20E-01 6.13E-01 5.61E-02 1.06E-01 9.90E-02 Isotepes Released Curies E degens 1-131 8.61E-05 1.68E-04 1.12 h0A 6.93E-Oh 5.52E-03 1.25E-02 1.91E-02 2.19E-02 I-133 3 3k b04 6.685-04 2.515-04 2.53E-03 1 33E-03 1.k1E-03 6.52E-03 2.hkE-02 1-135 2.21E-01 2.21E-01 2.21E-01 Pc.rticulates es-134 1.86E-05 k.15bO2 5.62E-Ok k.21E-02 b.22 E-02 2.73E-04 1.59E-02 2.21E-03 1.8kE-02 1.95E-02 Cs-137 k.1kbo5 BaLa-1k0 1.19E-03 1.65E-03 2.11E-Oh 1.62E-03 3.1kE-03 6.921-06 7.82E-03 1.kSE-02 Mn-54 1.0iE-Oh 1.0TE-04 4.5(F-Oh co-60 2.71E-04 3 76E-Ok 2.99E-05 1.2kE-04 1.95E-03 5 19E-05 2.80E-03 5 13E-03 Co-58 5.28E-05 Net Unidentified Beta 6.005-Ok 6.56E-Oh 8.52E-04 6.18bO3 k.SOE-03 T.03E-Ok 1 3kE-02 1 581-02 Noble Gases Ie-138 1.3kE+03 1.09E+03 1.10E+03 8.56E+02 1 52E+03 1.35E+03 T.26E+03 8.17t+03 Kr-87 1.12E+03 9.10E+02 8.10bo2 8.73E+02 1 93E+03 2 01E+03 7.65D03 8.55E+03 Kr-88 T.97E+02 T.50E+02 6.29E+02 6.89E+02 1.51 p03 1.23E+03 5 61E+03 6.16E+03 Kr-85a 3 33E+02 3 2TE+02 2 92E+02 3.kkE+02 7 5?E+02 k.79E+02 2.53D03 2.80E+03 Ie-135 1.3kE+03 1.b6E+03 1.19E+03 1.kOE-03 3.0TE+03 2.28E+03 1.0TE+0L 1.1890k Xe-133 5 71 bo2 5 95 b o2 6.hkp02 1.08E+03 1.62E+03 9 05E+02 5 k2E+03 5 79D03 Ie-lk3 p-9k Kr-93 Ze-1k1 Kr-92

.39E+00 Kr-91

<1

<1 Ie-Ik0 3.60E+00 3.53E+00 2.39E+00 1.36E+00 k

5.18n00 2.05E+01 2.3kr+01 Kr-90 k.13E+01 k.06 D01 2.76t+01 1.60E+01 5 11E+01 5.95E+01 2.36E+02 2 70E+02 Ie-139 6.13E+01 6.0?!+01 b.10E+01 2.kOD01 T.61E+01 8.82E+01 3.51 D 02 k.cloo?

Kr-87 8.95E+01 8.79E+01 6.11h01 3.86E+01 1.17p02 1.29E+02 5 23E+02 5.9fE+02 Ie-137 1.56 h02 1.53E+02 1.0TE+02 6.83E+01 2.0$E+02 2.2kE+02 9 13F+02 9.26E+02 Ie-135m 3 2kE+02 3 18E+02

2. 31E+02 1.69E+02 L. TOE +02 b.66E+02 1 98D03 2.2L D03 Kr-83 1.88E+02 1.85E+02 1.51D02 1.50E+02 3.55E+02 2.71E+02 1.30E+03 1.kSE+03 Ie-133m 1.6kt+01 1.62D01 2.35D01 k.51E+01 8.18E+01 2.37E+01 2.0TE+02 2.20E+02 Ie-111a

<1 41 2.83E+00 7.63E+00 1.25E+01

<1 2.60D01

3. TOE +01 Er-85

<1

<1 1.T5E+01

5. 36 E+01 8.53E+01

<1 1.59E+02 1.60E+02 N 13 1.72bO2 1.6Th02 1.61h02 1.58E+02 1.37E+02

1. 30bO2 9.25902 1.12E+03 70 P00R ORIGINAL

2 APPENN% B. TABEE 1 PEVTSED Consumers Power Company Big Rock Point Plant, Docket 50-155 Br 'loactive Liquid Releases E<x-Manth Units p rsy February March April May June Tctal Total Radioactivity Released (Except Tritium Dissolves Gases 1.59E-01 2 92E-01 1.15E+00 and Alpha)

Curies k.22E-01 7.6kE-02 1 98E-01 Volume of Waste Discharged Liters 7 69E+0h k.01E+0h 3 72E+0h 9.89E+0k L.06E+0L 2 9kE+05 Average Concertrathn of Waste Prior to Discharge pC1/ml 5.k9E-03 1.90E-03 5 32E-03 1.60E-03 7.18E-03 3 91E-03 Volume of Circulating Discharge

'eter Liters 8.k5E+09 T.78E+09 8.62E+09 8.50E+09 8.52E+09 8.18E+09 5.01Eu 0 Average Concentration Released (Except Tritium, Dissolved Gases t.nd Alpha) pCi/ml 5 00E-08 9.81E-09 2 30E-08 1.86E-08 3.5TE-08 2.30E-08 Maximum Concentration (Except Tritium. Dissolved Gases and Alpha) pC1/ml 9 05E-07 k.97E-07 5 77E-07 6 90E-07 3 9kE-07 9 05E-07 Psreent of Applicable Limits 5

5.87E+00 1.081#.00 k.17E+00 3 28E+00 7.T8E+00 L70E+00 Tritium Released Curies 7.69E-01 1.12E-01 k.8kE-01 2 37E+00 6.09E-01 L.3kE+00 Average Tritium Concentration Rnleased WCi/ml 9 10E-08 1.kkE-08 5.85E-08 2.76E-07 7.LkE-08 8.66E-08 Total Gross Alpha Released Curies 1.6kE-05 2.77E-05 5 72E-06 2 75E-05 k.12F-06 8.1kE-05 Average Alpha Cor. centration pCi/ml 1 9kE-12 3.56E-12 6.92E-13 3 23E-12 5.OkE-13 1.6kE-12 Irotopes Curies 6.25E-03 2.93E-03 9 92E-04 2.59E-02 Zn-65 1.57E-02 7.60E-03 I-1 31 7 60E-03 Fe-59 1.12E-03 1.1?E-03 Ce-13k 3 58E-02 1.09E-02 1 39E-02 7.BkE-03 2.61E-02 9 k5E-02 Co-137 1 93E-01 3.62E-02 b.21E-02 2.5LE-02 5 2kE-02 3.k9E-01 Bala-1k0 1 52E-03 1.52E-03 1.8hE-Ok 1.BkE-OL Co-58 Co-60 1.89E-02 k.25E-03 2.02E-02 3.L5E-02 2.06E-02 9.85E-02 Mn-5h 1.21E-02 7 0kE-Ok 9.k6E-03 3.6TE-03 1.SkE-03 2 78E-02 Br-89

1. 31E-03 k.L6E-05 9.89E-05 8.53E-05 1.5hE-03 Sr-90 2.k6E-Oh k.01E-05 1.15E-oh 1.LBE-oh 7.71E-05 6.26E-Ok Total Identified Released Radioactivity 2.86E-01 5 21E-02 9 21E-02 7.59E-02 1.02E-01 6.08E-01

)

Percent of Total Identifed I

6.78E+01 6.82E+01 k.65E+01 k.77E+01 3 50E+01 5 29E+01 71

(

P.00R.0RIGINAL

e f

Revised 2 15-76 APPENDIX P. TABLE 2 Consumers Power Company Big Rock Poir.t Plant. Docket 50-155 Radioactive Liguis Releases Sis-Month Yearly 311 July August Septemter Oeseber pavember Decerbar Total Tot al Total sadioactivity Released (Except Tritium, Dissolved Gases 2 57E-02 1.01E-01 1.38E-01 2.65 b01 1.k1E+00 and Alpha)

Curies Volume of Waste Discharged Liters 1.8kE+0h

1. k S E+0'.

6.77E+0h 1.01E+05 3 95E+05 Average Ccacentration of Waste Prior to Discharge bC1/a1 1.40E-03 6.97E-03 2.04E-03 2.62E-03 3.56E-03 Volume of Circulating Disebarge Water Liters 8.78E+09 8.78E+09 8.50E+09 8 79E+[9 8.k2E+09 8.7BE+09 5.21E+10 1.02E+11 Ave. age Concentration Pele(sed (Except Trititan. Dissolved Gases 3 02E-09 1.15E-08 1.6kE-08 5.09 E-09 1 35E-08 and Alpha) pC1/ml Maximus Concentration (Except Tritium. Dissolved Gasos and 5.67E-07 6.8kE-07 1.18E-06 1.18E-06 1.lfE-06 Alpha) pC1/a1 2.93E-01 2.2kE+00 2.29E+00 7.60E-01 2.20E+00 Percent of Applicable Limits 5

Tritium Released Curies 5.15E-03 1.7kh01 5.15E-01 6.9tI-01 5.Okt+00 Average Tritium Concentration Released pC1/a1 6.06E-10 1.98E-08 6.12E-08 1.33E-08 k.9kE-08 2.k8bo6 5 22E-06 6.89E-06 1.k(E-05 9.60E-05 Total Gross Alpha Beleased Curies 2.80E-13

9. k1E-13 Average Alpha Concentratico uC1/mi 2 92E-13 5.95E-13 8.18E-13 Isotopes Curies 1.23E-Ok 1.39E-03 1.51bO3 2.7kE-02 En-65 7.39E-05 7.68E-03 1-1 31 7 39E-05

.kTE-02 1.12E-03 Fe-59 C4-13k 6.393-03 1.09E-02 1.

3.20E-02 1.27E-01 3 48E-02 5.79E-02 L.cr7E-01 6.21E-03 1.69E-02 Cs-137 Bata-1ko 3.48E-Ok 1.11E-03 1.k(E-03 2.99E-03 7 33E-06 7.33E-06 1 92E-Ok Co-58 2.89E-03 8.kCE-03 2.96E-02 k.09bO2 1 39E-01 Co-60 3.0TE-03 6.k8E-03 k.30E-03 1 39E-02 k.17E-02 Mn-54 6.38E-04 6.77E-OL 1.30E-03 2.66E-03 Br-89 2.9hE-05 1.02E-04 7.k5E-05 2.06E-ok 8.30E-Ok Br-90 Total Identified seleased 1.86E-02 k.kOE-02 8.67bO2 1.k9E-01 7.58E-01 Radioactivity 7.2kE+01 k.36E+01 6.28E+01 5.62E+01 5.3SE+01 Per G-*

of Total Identified 5

Dissolved Nobis Cases Curies Ie-133 k.85E-03 k.85E-03 k.85E-03

2. 39E-03 2.39E-03 2.39E-03 Ie-135 72 P00R ORIGINAL

l APPENDIX C Off-Site Shipment of Radioactive haterial Shipnee.

Transfer Number Date From Transfer to Radioactive Material Disposition 370 1/ 9/75 DPR-6 Argonne National 1 Gal Liquid Vaste Sample (%.1 aci)

Analysis laboratory Exempt 371 1/15/75 DPR-6 NECo. Morehead, KY Solid Waste (3.kB6 C1)

Burial (16-NSF-1) 372 2/1L/75 Reesived 64 pCi Na-22 Source for Flux Wire Calibration 373 2/19/75 p.R-6 Isotope Products Lab Na-22 Source (6k WC1)

Return Burbank, CA (1509-59) 37h 3/ 6/75 DPR-6 Battelle Institute 8 Co 60 Rods (15k,000 C1)

Processing West Jefferson, OH (3k-0685k-05) 375 5/20/75 Received kk.2 pCi Na-22 Source for Flux Wire Calibratica 376 5/2L/75 Received 0.1 mci Each of Ra-226 and Se-75 for Instrument Cc libration 377 o/11/75 DPR-6 NEco,Ibrehead, KY Spent Resin (%70 C1)

Burial (16-NSF-1) 378 6/13/T5 DPR-6 NEco, Morehead, KY Spent Resin (96 C1)

Burial (16-NSF-1) 379 6/16/75 DPR-6 NEco, Morehead, KY Spent Resin (8k C1)

Burial (16-NSF-1) 380 6/17/75 DPR-6 NECo Morehead, KY Spent Resin (106 C1)

Burial (16-NSF-1) 381 6/18/75 DPR-6 NECo, Morehead, KY Spent Resin (125 C1)

Burial (16-NSF-1) 382 6/19/75 DPR-6 NECo. Morehead, KY Spend Resin (Sk C1)

Burial (16-NSF-1) 383 6/25/75 DPR-6 NECo Morehead, KY Spent Resin (50 C1)

Burial (16-NSF-1) 38k 6/2k/75 DPR-6 NECo, Morehead, KY Spent Resin (30 Ci)

Burial (16-NSF-1) 385 6/25/75 DPR-6 NECo Morehead, KY Spent Resin (90 C1)

Burial (16-NSF-1) 386 6/26/75 DPR-6 NECo. Morehead, KY Spent Resin (75 C1)

Burial (16-NSF-1) 387 6/26/75 DFB-6 NECo. Morehead, KY Spent Resin (124 Ci)

Burial (16-NSF-1) 388 6/27/75 DPR-6 NECo, Marehead, KY Spent Resin (100 C1)

Burial (16-NSF-1) 369 10/lk/75 DPR-6 NECo. Morehead, KY 153-55 Gal Drums (218 mC1)

Burial (16-NSF-1) 390 10/23/75 DPR-6 Argonne National b Irradiated Fuel Rods Examination lats (Exempt)

(1,976 01) 391 11/ 6/75 DPR-6 NECo, Morehead, KY Solid Waste k.16 Ci Burial (16-NSF-1) 392 11/13/75 DPR-6 NECo, Morehead, KY Solid Waste 3.2 Ci Burial (16-NSF-1)

P00R ORIGINAL

Appendix D ENVIRONMENTAL THEIE0 LUMINESCENT COSIMETERS BIG ROCK POINT 1970 Series G Series O Standard Standard Deviation Deviation Average (2 Sigma)

Average (2 Sigma)

February 16.7 3 00 1

March 13 5 1 54 7.7 2.04 April 16.2 3 38 May 18.5 1 92 95 0.62 June 18.5 3.88 July 19 9 1 92 10.8 0.84 August 15 9 3 16 September 16.8 2.02 9.o 1.So october 18.4 3 72 November 16.3 2.42 8.6 1.36 December 17 1 3 34 Weighted Average 16.8 2.40 9.8 o.98 74 m

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APPENDIX D High, Low and Average Concentrations For Highest Average Sampling Location January 1, 1975 to June 30, 1975 M

Type of Analysis Units Location High Low Average Air Gross pCi/m3 TR 0.25 0.02 0.08 3

I-131 pCi/m ALL.

0.03 0.2 0.2 Lake Water Gross Beta pCi/l BR ST IWo 43 4.3 12.0 H-3**

PCi/l BR ST IMO 740 300 540 Well Water Grcss Beta BR ST W 7.6 0.3 2.5

TID Dose mR/mo C

6.1 29 51

In excess of contml donimeters.

- Includes January through November only.

1 i

I

i l

Appendix D AquaticBiota,pCi/g Discharge Fish Nuclide Suckers Alewives Trout 6/75 10/75 6/75 10/75 6/75 10/75 Gross Beta 4.8 (5.4 4.1 Im 1.8

< 2.7 Mn-54

<o.02

<o.01

<o.02

<o.02 Fe-59

<o.06

< o.09

<o.03

< o.08 ce-58

<o.03

<c.03

<o.01

<o.02

<o.03 co-60

<o.03

<o.02

< o.01

< o.02

<o.02 zn-65 (o.oS

<o.05

<o.03

< o.04

<o.05 Zr-93

<o.06

<o.07

<o.03

< o.04

<o.02 Nb-95

< o.02

<0 02

<o.01

<o.02

<o.02 I-131

< o.02

<2

< o.02

<o. 02

<1 cs-134

< o.02

< o. 02

<o.01

<o.02

< o.02 cs-137

<o.10

<o.14

< o.14

< o.18

<o.18 BaLa-140 (o.04

<o.03

<o.02

< o.03

< o.04 K-40 1.1 2.6 2.61 1.4 2.2 0.02 (o.03 Th-228 sr-89

< o.007 NA v.007

<o.004 NA sr-90

< o.004 NA o.025 0.020 NA NS (No Sample)

NA (Not Available)

Appendix D Aquatic Biota, pCi/g Crayfish Nuclide Discharge dmiSouth hMiNorth Mt McSuba Nine Mile Pt 6/75 lo/75 6/75 10/75 6/75 10/75 6/75 10/75 6/75 10/75 Gross Beta 37 6.6 30 4.4 2.2 5.2 0.5 35 32 4.2 Mn-54 0.11 0.08

< o.c3

< o.04

<o.03 0.06

<o.03

<o.04

<o.o3

<o.1 Fe-59

<o.o9

< o.08

< 0.09

< o.1

< o.04

< o.1

<o.08

< o.17

<o.08

<o.36 co-58

<o.04

< o.02

<o.04

<o. 04

< o.03

< o.06

< o.04

< o.06

< o.03

<o.13 co-60 0.28 0.43

<o.04

< o. 04 0.11 0.29

<o.04

<o.oh

<o.04 0.13 zn-65

< o.01 0.11

<o.o8

< o.08

<o.07

< o. 02

<o.08

< o.10

<o.07

<o.23 zr-95

< o. 09

<0.06

<o.08

<o. 07

<o.07

< o. 08

<o.08

< o.08

<o.08

<o.02 Nb-95

< 0.04

<0.02

< o.04

<0.03

<o.03

< o.04 40.04

<o. 03 0.04

< o.08 I-131

< o.o8

<o.3

<o.1

< o.03

<o. 07'

< o.03

< o.09

<o.03

<o.09

<o.07 cs-134 0.13 0.11

< o.ok o.06 0.10

<o.04

<o.03

<o.o4

<o.03

<o.o9 cs-137 o.68 o.4 0.22 0 30 0.49 0 36 0.17 0.14 0.17 o.4 BaLa-lho

< o.08

< o. 04

<o. 06

<o.06

< o.06

<o.07

< o.06

<o.o?

<o.06

<o.2 K-ho 2.2 2.2 1.4 2.7 1.6 2.2 19 32 19 8.0 sr-89 NA NA NA NA NA NA (o.ol NA

< o.008 Sr-90 0.69 NA o.49 NA o.83 NA o.44 NA

<o.004 NA (Not Available)

Appendix D AquaticBiota,pCi/g Periphyton Nuclide Discherge hMiSouth

{MiNorth Mt McSuba Nine Mile Pt 6/75 l

10/75 6/75 10/75 6/75 10/75 6/75 10/75 6/75 10/75 Gross Beta 11.4 14 3.7 2.8 6.3 79 NS 53 NS 3.8 Mn-54 1.21 5.1

<o.3

< o.09 0 38 0 90

< 0.02 0.09 Fe-59

< o.11

< o.04

< o. 06

< 0.03

< o.05

< o.2

< o.06

<o.08 co-58

< o.08

< o.1

< o.3

<o.11

< o.03

<o.11

<o.02

< o.03 co-6o 37 15.0 0.8 o.58 1.41 2.8

<o.02 0.19 zn-65 0.8 2.6

< o.6

<o.22

< 0.08 0.29

< o.04

< o.05 Zr-95 o.24

<o.27 09

<0.27 0.15

<o.2

<o.03

<o.07 Nb-95 0.19

< o.07 1.8

< o.07 0.18 0.05

<o.02

<o.02 I-131

<o.25

<6

<1

<8

< o.06

<4

< o.01

<2 cs-134 1.12 0.69 1.0 0.18 0.42 0.26

< o.02 0.03 ca-137 4.5 2.5 6.7 19 2.74 2.0 0.18 0.23 BaIa-lho

<o.15

< o.12 (o.5

<o.13

< o.07

<o.08

< 0. 02

< o. 03 K-40 52 5.6 4.0 16.o 25 2.6 6.2 4.9

'Th-228 0 38 0.10 1.7 0.73 0.09 0.10 0.33 0.11 Ra-226 o.18 0.86 0.55 0.16 0.07 0.22 0.11 o.2 o.48 5.4 1.7 1.1 Be-7 0.4 Ru-106 Sr-89

<0.01 UA

< o.ol NA

<o.01 NA NA NA Sr-90

< o. 004 NA o.097 NA o.079 NA NA NA ns (No Samnle)

NA (Not Availablel

Append 14 D AquaticEtato,pCi/g Algae Nuclide Discharge dMiSouth dMiNorth Mt McSuba Nine Mile Pt 6/75 1o/75 6/75 10/75 6/75 10/75 6/75 10/75 6/75 10/75 Gross Beta 17.6 NS NS 2.6 NS NS 5.4 2.0 2.1 NS Mn-54 1.25

< o.02

<o.02

< o.02 Fe-59

< o.12

<o.08

<o.ok

<o.06

<o.04 co-58 o.15

<o.03

<o.o2

<0.02

< o.02 co-60 4.81

< o.02

< o. 02

<o.02

<o. 02 zn-65 0.8

<o.05

<o.04

< o.04

< o.04 Zr-95 0.24

<o.07 0 35

<o.o3 o.21 cn Nb-95 0.17 0.04 0.26

<o.02 0.16 I-13 1

<o.2

<2

< o.08

<o.01

<o.07 cs-134 1.14

< o. 02

<o.02,

< o.02

<o.02 cs-137 4.84

<0.02

.07 0.18 0.08 1

BaLa-lho

<0.2 (0.03

<0. 04

<o.02

<0.04 K-40 6.5 2.4 1.2 6.2 2.5 0 33 Th-228 0.13 0.C,

0.22 Ba-226 o.2 0.02 0.09 o.2 0.52 Be-7 Ru-106 Sr-89

<o.007 NA

<o.03 NA

< 0.01 Sr-90

<o.004 NA (o.ook NA o.ok

Apprndix D AquaticBiota,pCi/g Shore Minnows Nuclide leischarge hMiSouth Mi North Mt McSuba Nine Mule Pt 6/75 l

10/75 6/75 10/75 6/75 10/75 6/75 10/75 6/75 10/75 Gross Beta NS NS NS NS 4.4 39 2.2 3.2 NS 2.0 Mn-54

<o.02

< o.09

< o.03

<o.08

<o.2

< o.4

<o.08

<o.04 (0.6

< 0.06 Fe-59 co-58

< o.03

< 0.1

< o.ok

< o.1

<o.2 co-60

<o.03

< o.1

< o.oh

< o.08

< o.1 d0.06 03

<o.08

< o.2

<o.4 An-65

<o.06

<o.3

< o.08

< o.3

<o.5 Zr-95 8

Nb-95

< o.02

<o.07

<0.03

< o.06

< o.o1

<a.08

<6

< o.08

<6

<6 I-131 cs-134 0.08 0.15

<o.04

<o.o7

< 0.1 0.42 0.82

< o.04 0.14 0.02 cs-157 BaLa-lho

< 0.05

< 0.1

<o.06

< o.1

< o.02 1.4 6.3 2.4 5

0.8 K-40

< o.2 0.11 Th-228 na-226 Be-7 Ru-106 Sr-89

<o.005 NA NA NA NA Sr-90 0.02 NA o.08 NA NA

APPENDIX D Sampling and Analysis Summary January 1, 1975 to December 31, 1975 No of Samples Frequency of Meditan Description Location Collected Type of Analysis Analysis Air Continuous at All 37G Gross Beta, I-131 Weekly Approximately r

1 cfb Iake Water 1 Gal ST, CH 36 Gross Beta Monthly Well Water 1 Gal Grab ST 12 Gross Beta Monthly Gamma Dose Continuous All 142 TID Dose Monthly Aquatic Biota Grab ST, NM, 33 Gross Beta, Isotopic Semiannual Mt McSauba l

l l

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