ML20030A498

Semiannual Operating Rept,Jul-Dec 1972

ML20030A498
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	12/31/1972
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090770
Download: ML20030A498 (62)

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CONSUbERS POWER COMPANY BIG ROCK POINT PIANT SEMIANNUAL OPERATIONS REPORT NO 17 JULY 1 - DECEMBER 31, 1972 N

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i CONSLMERS PCWER COMPA?iY Docket No 50-155 License No DPR-6 17TH SIMIANNUAL REPORT OF OPEPATIONS OF BIG ROCK POIIiT NUCLEAR FIANT July 1, l$r/2 - December 31, 1972 I.

INTRODUCTION - SEGANNUAL OPEPATING REPORT The plant was base loaded at 63 se (gross) during this report

- period to limit fuel cladding heat flux. The off-gas release rate on July 1 was approximately 1,000 pCi/second.

II.

OPEPATIONS SUWJGY A.

Changes in Plant Design No changes in plant design occurred during this repcrt period.

B.

Perfomance Characteristics A load rejection test was conducted on July 6 at a plant output

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of-63 MWe(g). The. control circuits and bypass valve operated satisfac-torily; however, the feed-water supply could not be maintained and re-

. sultad in a' low drum level scram. For details, please refer to Section VI, " Changes, Tests and Experiments." The plant was returned to service on July 7.

On July' 27, control rod drive E-1 inserted to 00 while work was perfomed on a scram valve solenoid which resulted in a power reduction of 10 MWe.- The scram valves were inadvertently cpened, causing the inser-

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' tion. The control' rod was returned to Position 23 within a few minutes and power restored to 63 MWe (gross).

The plant was' removed from service on July 24 for a scheduled-outage to repack the turbine main steam bypass valve. During this period, the circulating water intake crib structure was inspected with no problems being observed.. In addition, the failure of both neutron monitoring start-up channels -delayed plant start-up (for details, see Section V, " Safety-Related Maintenance,": Item A.4). The plant returned to service on July 30

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and was at 63 MWe (gross)- on August-1.

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Degradation of the No 1 reactor recirculation pu=p seal was first observed on August 24 (this seal was replaced in July if/2). This degradation was the result of a restriction in the outer seal breakdcwn orifice.

During routine quarterly testing on August 31, 1972, and again on November 23, 1972, the containment inside isolation valve on the clean-Details of these up and fuel pool _ drain system was found to be inoperable.

failures are reported to the AEC by letters from Const=ers Power Co=pany dated September 6,1972 and December 20, Igr/2. This type of failure has occurred previously (see 15th Semiannual Report) and is being studied by the valve manufacturer and Consumers Power Ccmpany. Special procedures for weekly testing of these valves have been instituted.

On September 14,1Cff2, during a routine test, the emergency diesel generator failed to come up to voltage. Investigation revealed that the exciter ar=ature had deteriorated and would require off-site repair.

During this period of repair, a portable generator was obtained and con-nected to supply power to the emergency 2B bus. This system was restored to nor=al on September 24. Details of this failure were reported to the AEC by letter dated September 18,1Cf(2.

The unit was forced out of service on September 30 to repair a steam leak on the high-pressure extraction line of the turbine. The plant returned to service on October 2.

On September 30 and again on Nove=ber 10, the turbine bypass iso-lation valve failed to close. The initial investigation of this problem indicated that the probable cause of the failure was that the electrical lead interference caused.a mechanical bind on the L5ritorque centrol. Fol-lowing the secorA failure, the problem wa.s reevaluated and plans made to adjust the upper limit switch to prevent backseating of the valve to de-termine if this resolved the problem.

The clean-up system pu=p failed on November 6 due to pu=p bearing failure.. The spare pump was installed and the clean-up system returned to service on November'8. Plant load was reduced to 10 MWe (gross) on both

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dates to facilitate system isolaticn and the return to service of the clean-up system.

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On November 10, the plant was removed frau service on a sched-uled outage to perform the control red drive required six-month test as required by the Technical Specifications.

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Additional work perfomed during the outage included:

1.

The 0-rings on control rod drives B-1 and B-3 were replaced because of leakage from the drive flanges.

2.

A new cooling water pump was installed on the emergency diesel generator. This new pump was satisfactorily flow-tested.

During start-up on November 12, the reactor scra=med on short period because of a higln notch worth in sequence during withdrawal of con-trol rod B-5 A new control rod withdrawal sequence was developed to mini-mize this operating difficulty. The plant was returned to service on November 13 The No 1 condenser circulating water pump developed excessive vibration and was removed from service on November 15 Examination of the pu=p indicated that the lower bearing had failed, causing damage to other

. pump part.s. The pu=p was repaired and returned to service on November 29 The 0-ring on control rod drive C-5 had deteriorated to the point where excessive leakage forced the plant out of service on November 23 New 0-rings were installed and the plant returned to service on November 25 On December 16. the plant was removed from service on a scheduled outage to repair a leak in a feed-water line blank flange. The unit was i

returned to service on December 17 and was at 63 We (gross) on December 18.

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_ The plant o&-gas reached 26,000 pCi/s on December 30 and plant load was reduced to 52 We:(gross) to reduce the off-gas release rate.

_ Off-gas at 52 We (gross) averaged about 15,000 pCi/second.

C. ' Changes in Procedures Which Were Necessitated by A and B or Which Otherwise Were Required To Improve tne Safety of Facility Operations The following lists the procedural changes made with respect to

. plant operations:

A2 51 - Describes memos-to. Operating Personnel pertaining to plant operations and specifies where they are to be maintained.

A2.5.2 - Specifies the numbering and control system of menos per-taining to plant operations. -

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Bl.l.91 - Limits rate of power ascents following an outage.

Bl.6 - Defines action to be taken following a suspected LOCA.

I B26.3 2.4 - Describes elimination of the automatic timer to shut the electric fire pump down manually.

B28.3.6.5 - specifies shutdown of emergency diesel generator through manual operation of the governor control.

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- E2.1.5 - Limits entry of the recirculating pump room while at i

l power.

E31.4 - Describes the proper procedures to be used for personal decontamination.

f E3 2.1 - Describes the alarm set points for continuous air monitors.

a D.

Results of Surveillance Test and Inspection Required by Technical Specifications

The following' listing shows the systems tested, the required test I-frequency, the dates tested during this report period, and the results of the test (s):

- Containment Isolation System: Containment isolation valve controls, and instrumentation.

Required Frequency: Quarterly.

Test Dates:. July 6 and September 30.

Results:.The automatic controls and instrumentation for eight of the-

-nine isolation valves vere checked and found'to function properly. One. valve (main steam drain, MO-7065) is maintained

.in the closed position,' de-energized and not used. Therefore,

testing the automatic centrols of this valve is considered unneeessary.

System: ' Containment sphere component leakage rate -test.

. Required Frequency: Six months or less.

Test Dates: ' October 11..

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Results: The' containment sphere component leakage rate test was per-formed using' air at 20 psig.. The results of'this' test'showed a.' total le'akage.of 40% of the allowable ' limit. Adjustment of.a packing leakLon one:offthe. valves reduced the total leakage to h

35% or the,anowable

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Systen: Centainnent sphere isciatien trip circuits.

Required Frequent': Daring each =ajor refueling shutd v. but not less frequently than ence every twelve ncnths.

Test Dates: No test required during this report peric>d.

Results: N ne.

Syster: Isolatien valve lee's and c;erability test.

Required Frecuencv: Twelve ncnths or less.

Test Dates: No test required during this report pericd.

Results: Ncne.

Systen: Centainnent sphere penetratien inspecticn.

Required Frequency: Twelve nenths er less.

Test Dates: No test required during this repcrt period.

Results: Nene.

Sgsten: Centainment sphere integrated leak rate test.

Required Frequency: Every tv: years.

Test Dates: No test required during this repcrt period.

Results: Nene.

Centrol Rod Drives and Associated Equiptent Syste : Reactor safety systen scra circuits (not requiring plant shutdown to test).

Recuired Frequenev: One conth or less.

Test Dates: June 18, July 6, July 30, August 30, Septenber 29, October 27, November 2L and Dacember 16.

Results: The reacter safety syste= sas tested using the switches pro-vided to sirulate sensor trips. All channel trips occurred as desie,.ed. In additien, the neutren =enitoring p0ver range m

and interredis.te range channels were tested for trip setting.

All of these tests shoved the trip settings to be within the 120 +., 2% cf pcver and the 10-secend periods set +,ing, respec-tively, with the following exception: On August 30, pcver range channel No 1 test showed a trip setting of 11% cf power. This setting was corrected.

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I System: Control rod perfomance - run.

Required Frequency: Each major refueling and at least once ever/ six months during period of power operation.

Test Dates: November 12.

Results: The contrcl rod drive continuous withdrawal and insertion s

test, including withdrawsl timing, was perfomed for each drive. This test is perfomed during reactor shutdown fol-lowing cocpletion of other drive perfomance tests and adjust-l ments and represents the results of the final timing of each drive. The results of this test showed all drives to be

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cperatir.g satisfactorily with scst withdrawal times at 36 l

I seconds. No withdrawal time was less than 23 seconds.

System: Control rod pe-fomance - jog.

l Required Frequency: Each major refueling and at least once every six l

months iuring period of power operation.

Test Da b ?: November 12.

i Results: The control rod drive latching test was performed for each drive. This ~ test showed satisfactory latching of all drives.

System: Control rod performance - scram.

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

Test Dates: November 10.

Results: The control rod scram time test was perfomed for each drive.

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t-The test included time from system-trip to 1007o of insertion at a reactor water te=perature of about 150 F.

The results cf this test were satisfactory for all drives with scram times

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ranging between 1.04 and 1.46 seconds.

System: Reactoi safety system scram cire".its- (requiring plant shutdown to test).

Required Frequency: During each major' refueling shutdown bat not less frequently than once every twelve months.

ETest Dates:. No test required during this report period.

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Results: ~None.

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7 System: Reactor safety system response time (requiring plant shutdown to test).

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

Test Dates: No test required during this report period.

Results: None.

System: Control rod withdrawal permissive interlocks function.

Required Frequency: Twelve months or less; the r fueling interlocks will be tested prior to each major refueling.

Test Dates: No test required during this report period.

Results: None.

System: Centrol rod drive friction tests.

Required Frequency: During each major refueling out not less frequently than once a year.

Test Dates: No tent required during this report period.

Results: None.

Emergency Cooling System: Core spray system check valves.

Required Frequency: Twelve months or leas.

Test Dates: No test required during this report period.

Results: None.

System: Post-incident spray system automatic control operation.

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

Test Dates: No test required during this report peried.

Resu hs: None.

System: Reactor emergency cooling system trip circuits.

Required Frequency: Twelve months or less.

Test Dates: No test required during this report period.

Resu'.ts: None.

Miscelictneous Systems Systm: Shutdown margin test.

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Required Frequency: After each refueling and certain core component changes and if system is cooled and 3!+,000 IWdt have been generated.

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'i Test Dates: November 12.

Results: The shutdown margin verification test was performed using two adjacent fully withdrawn control rod blades and a third adjacent blade withdrawal to a position known to contribute 1

This test was performed on selected at least 0.003 keff.

maximum worth rods and demonstrated. a satisfactory shutdown j

margin.

System: Nil ductivity transition tengerature calculation.

Required Frequency: At least once each year.

Test Dates: No calculation required during this report period.

Results: None.

System: In-service primary system inspection.

' Required Frequency: A continuing program being conducted every major refueling outage.

Test Dates: No test required during this report period.

Results: 'Ncae.

1 System: Moderator temperature coefficient.

Required Frequency: Following each major refueling.

Test Dates: No test required during this report period.

Results: None.

-System: Suberiticality checks.

Required Frequency: During ' core alterations which increase reactivity.

Test Dates: No test required during this report period.

- Results: None.

System:. Refueling operation controls.

Required Frequency: Each major refueling.

Test Dates: No test required during this report period.

Results: Ront..

System: Reactor refueling safety system sensors and trip devices.

Required-Frequency: Each major refueling.

Test Dates: No test required during this report period.

Results: None.

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Poison Systems System: Liquid poison system firing circuit test.

Required Frequency: Two months or less.

Test Dates: August 28, October 30 and December 3 Results: The liquid poison system circuits monitoring the poison injection valves circuits "A" and "B" were checked by opening the isolation breaker for each circuit. The monitors func-tiened properly on each test.

System: Explosive valve from equalizing line.

Required Frequency: Twelve months or less.

Test Dates: No test required during this report period.

Results: None.

System: Explosive valve from nonequalizing lines.

Required Frequency: Twelve months or less.

Test Dates: No test required during this. report period.

Results: None.

Radiation Monitoring Syste=: Air ejector off-gas monitor system.

Required Frequency: One month or less.

-Test Dates: July 31, August 25, September 29, october 26, November 21 and December 28.

Results: The air ejector off-gas monitoring system logarithmic monitoring.

instrument was checked using the instrument calibration test points at 1,10 and 1 x 1G' units. The checks showed the cali-

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bration to be' satisfactory (within 20fo).

The automatic. closure function of the isolation valve ti=cr

. as checked. The test showed the timer calibration t,o be w

satisfactory (within-3% of the maximum timer setting) and the isolation valve,to close as specified.

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

-system.

Required Frequency': One month or'less.

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-Test Dates:. July 31, Angust 25, September 26, October 27, November 21 and December'28.

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Results: The stack-gas =enitoring system was checked using the built-l in calibration source (Cs-137). The instrument check chewed the calibration to be satisfactory, resulting in the alarm point occurring within the specified 0.1 curie per secend release rate.

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 the iodine filter are reported in terms of curies re-leased in Appendix A of this report.

System: Calibration of emergency condenser vent monitors.

Recuired Frequency: One month or less.

Test Dates: July 31, August 25, Septerle 29, October 26, november 21 and December 29 Results: The emergency condenser vent monitors are cheched by comparing I'

with a calibrated portable instrument. The checks showed-the vent monitor calibration to be satisfactory with all menitor

- checks within 1 5% (of full scale).

System: -Calibration of canal liquid process monitor.

Required Frequency: One month or less.

. Test Dates: July 31, August 25, September 26, October 27, November 21-

- and December 28.

Results: The calibration of the " radioactive vaste system effluent to.

' canal" monitor is a comparative calibration used to' demonstrate operations of.the monitor,and to detect gross calibratien

' changes'. The results of these monthly calibrations showed that the monitor was ntnctioning properly and that no gross monitor-drift had occurred since the original calibration which

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' utilisied certified standards.

' System: Canal liquid cellection sample..

? Required Frequency: Daily, f

Test Dates: The' analysis was. conducted daily.

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11 Results: The results of the daily analysis of the 2k-hour canal collection sa=ples are reported in Appendix 3 of this report.

System: Calibration of area monitors.

Required Frequency: One month or less.

Test Dates: July 31, August 25, Septe=ber 29, october 26, nove ber 21 and December 29 Results: The area monitor calibraticns are checked by conparing read-ings with a calibrated portable instru=ent. The checks showed the area monitor calibration to be satisfactory with rest coni-tors within 15% (of full scale) and all sonitor calibratiens within t 10%.

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

Required Frequenev: Three =onths or less.

Test Dates: July 31, August 25, September 28, October 27, November 21 and Dece=ber 28.

Results: The calibration of the liquid process tenitors (except the canal monitor which is reported separately) is a co=parative calibration used to de=onstrate operation of the nonitor and to detect gross calibration changes. 'The results of these monthly calibrations showed that the monitors were functioning properly and that no gross drift had occurred since the origi-nal calibration which utilized certified standards.

E.

The Result of Any Periodic Contain=ent Leak Rate Test Perforned During the Reporting Period No integrated containment leak rate test was performed during the report period. -

F. ' Chances, Test and Experiment Requiring Authorization From the CeM ssion During this report pariod, the following Technical Specifica'. ions changes'were authorized by the Comission:

Change No 31 - This propose'd change was later ccebined with No 3k.

Change No 32 - This changed the specifications to include prhry

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_ coolant' leakage W1 ts. -

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l 12 Chance No 33 - This change concerned the testing of the explosive-actuated valves of the liquid poisen system.

Chance No 3h - This changed the specifications to permit operation with Reload G and Nuclear Fuel Services Demonstration Assemblies.

G.

Changes in Plant Operating Organization Involving Key Supervisory Personnel During this report period, one change in key supervisory personnel occurred. On Novenber 1,15f72, R. W. Voll replaced D. A. Bixel as Reactor Engineer.

Mr. Voll received his BS Degree in Physics from Aquinas College and a Masters in Nuclear Engineering frc= the University of Illinois. His working experience consists of two years as Assistant to Reactor Engineer at Big Rock and one and one half years as Acting Test Engineer at the Palisades Plant. While at Big Rock, Mr. Voll also received his operating license.

Mr. bixel is now serving in the capacity of Senior General Engi-neer at the Big Rock Plant.

III. POWER GENERATION Report Total Period To Date (i) Themal. Power Generated (MWh )

812,500 10,799,5771 t

(ii) Gross Electric Power Generated (MWheg) 258,495 3,439,946 (iii) Net Electric Power Generated (MWhe) 2hh,523 9 3,240,550.6 (iv) Hours Critical (h) 4230.0 61,6h9 0 (v) Hours Generator On Line (h) h201.6 59,887.0

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IV.

SHUTDOWNS During this report period, six outages occurred. A description of each outage, listed in chronological order, follows below:

OUTAGE REPORT Type of Outage - Scheduled (8-72)

Length of Outage - 13 h, 2 min Unit Off Line - 1502 h, 7/6/72 Unit on Line - 0404 h, 7/7/72 A scheduled load rejection test at 63 MWe was conducted on July 6, 1972. Performance of the main steam bypas: valve at this load was satis-factory; however, the reactor scrammed due to low steam drum water level.

For details, please refer to Section VI, " Changes, Tests and Experiments."

The unit remained in the hot shutdown condition for the duration of the outage.

OUTAGE REPORT Type.of Outage - Scheduled (9-72)

Length of Gutage - 29 h, 32 min Unit Off Line - 0048 h, 7/29/72 UnitonLine-0620h,7/30/72 The plant was removed from service on a scheduled outage to repack the turbine main steam bypass valve because of a steam leak.

Following repairs, start-up was delayed due to instrumentation problems that developed. Failure of both neutron monitoring start-up channels occurred due to mechanical damage to the chamber and cable com-ponents caused by high ambient temperatures'iit the chamber guide tubes.

For details, please refer to Section V,l " Safety-Related Maintenance,"

Item 4.

The unit was. maintained in the hot shutdown condition during the course of the outage.

OUTAGE REPORT Type of Outage - Forced (10-72)

Length of Outage - 39 h, 59 min UnitOffLine-1526h,'9/30/72' UnitonLine-0725h,10/2/72-The plant was forced out o'f service when a steam leak developed in the '10" turbine extraction' line to the high-pressure feed-water heater.

Cause of the leak was due to internal: erosion on the outside radius on an elbow.. This elbow will be scheduled for replacement during the next

(;

' refueling outage.

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15 t

The method of shutting down was a controlled, deliberate shut-down and the unit was naintained in the hot shutdown condition during the outage.

OUI' AGE REPORT Type of Outage - Scheduled (11' 72)

Length of Outage - 72 h, 37 =in Unit Off Line - 0220 h,11/10/72 Unit on Line - 0257 h, 11/13/72 The plant was removed frc= swrice en a scheduled outage to perform control red drive tests as required by the cperating license.

The unit's status during the outage was cold shutdown.

OUTAGE REPORT Type of Outage - Forced (12-72)

Length of Outage - 32 h, 49 min Unit Off Line - 20k7 h,11/23/72 Unit en Line - 05L6 h,11/25/72

- The plant was re=cved from service en a forced outage because of excessive cooling water leakage on C-5 centrol red drive flange.

- Fiange 0-rings were replaced on the drive because of the exces-sive leakage. For_ details, please refer to Secticn V,'" Safety-Related Maintenance," Item C.3 The method of shutting down was a controlled, deliberate shut-down and the unit's status during the outage was cold shutdown.

OUTAGE REPORT

?jpe of Outage '- Scheduled (13-72).-

Length of Outage - 26 h,13 min

. Unit Off Line - 0105 h,12/16/72 Unit on Line - 0318 h, 12/17/72

~ The plant was removed frcm service on a scheduled outage to effect repairs to a leak on a " blank flange" in the feed-water header

,just below the steam drun.

' Inspection revealed a cut in the metal flange =ating tongue surface. Repairs were effected by filling the cut with Devcon stainles's steel filler and ins +&tien of a new Flexitallic gasket.

The unit's status during the outage was cold shutdcwn.

b S'

16 V.

SAFE'"Y-PHATED MAINTENANCE Note: Dates contained in this section generally refer to weekly period when the =aintenance was perfor=ed.

A.

Feactor Prctection and Centrol Syste: Instru=entation 1.

Neutren !4cnitoring Channel No 1 7/27/72 - The dual high-voltage power supply fer this chan-n.

nel was replaced with a spare unit folloving loss of regulation of the co=pensated ion 'chanher polarizing potential. The voltage increased to greater than -1500 volts (nor= ally -300 volts); however, no change in

+

picoa =mter output occurred as the cha=ber is fully saturated at approx-i=ately -100 volts during power operation.

Eench testing of the failed unit revealed a defective elee-

-tron tube in the series regulator control circuit. This type of failure is considered to be within the design li=itation of the equip =ent.

b.

8/31/72 - The cha=ter and coaxial cables (frc= the chamber to the cha=ber drive head) were' replaced in this channel due to trip point ' interference end erratic output.

During the =cnthly reactor protecticn syste= testing, it was noticed that the trip points on this che=nel vere at 102% (alar =) and 115% (trip) instead of the nor=al 105% and 120'4. Also, following this trip test, the picoan=eter output began.inereasing for no apparent rea-son. The picoa=neter was replaced with the spare unit. This, hcVever, did not correct the proble=,

he picon--eter which had been re=oved was bench-tested and reinstalled. - Tests of the cha=ber and cables indicated that lov re-

.sistance in the syste=.vas the proble=, resulting in indueed a-c pickup en the signal cable, and t::e cha=ber and cable vere replaced (the cha=-

ber replacer.ent was to expedite repair' and not.due to chamber failure).

- The high-voltage-power supply for this channel was also replaced prior.to returning theEsyste= to norwal. Failure of this unit-was traced to a dirty contact in the time delay relay schere which delays

' full output voltage for 30 seconds. after equipment turnon. Nor=al polar-

~h izing. potential-for thit, channel is -600 volts and the supply is li=ited to -500 volts-during varm-up time (during power operation, 'the cha=ber is

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saturated at les's tha: -100 volts).

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I 17 c.

9/7/72 - The dual high-voltage pow supply in this channel was replaced with a spare unit following loss of positive (com-pensating) voltage output.

Loss of the positive voltage has negligible effect on the chamber output when the reactor is at power. Repairs to the failed supply consisted of replacement of one shorted filter capa-citor and an overheated resistor.. This type of failure is considered to be within the design limitations of the equipment.

d.

10/5/72 - The picca==eter in this channel was repaired

'to correct minor inaccuracies during range-switching below the 0.h%

power range.

Initial observation of the proble= indicated that the feedback resistance in the range switch had changed value. The range switch was removed and bench-tested. However, testing of the switch indicated that all. ranges were within calibration and the switch was performing properly.. The picca==eter vas removed and bench testing revealed that a shift in balance had occurred.'

The unit could be rebalanced properly but was unstable and susceptible'to vibration. Replacement of the vibrating capacitor in the a-c a=plifier corrected the problem; all other parts tested nornal.

The unit' was' returned to service following "in-panel" balancing.

The Technical Specifications and plant design provide for re-moval of one power range flux monitor fro = service without compromising safety.

-2.

Neutron Monitoring Channel No 3 a.

10/5/T2 - The inverter power ' supply for this channel

- tripped off during a severe electrical sotre on September 28. The unit-

.vas returned to service'following testing. As the station battery was

- on overcharge' prior to the trip,'it can only be' assu=ed that lightning caused an overvoltage on the h80 -V a-c battery charger supply'resulting

~

in excess -inverter input current. - The inverter has performed properly since its return to service.-

,b. '12/21/72 - The picoammeter.in this channel was replaced

.[

vith'the spare unit to eliminate the minor variations.(approximately 2%)

in the recorder trace which had'recently been. occurring..

n

18 i

The inverter had previously been suspect, as the pico-ammeter behaved normally when placed on the alternate power source. An additional output level recorder was utilized to eliminate the picoa= meter recorder as the problem and the voltage output level of the inverter was monitored for any simultaneous excursions of the inverter output voltage.

Also, the output voltage level of the inverter was varied in an effort to induce variations in the picoarmeter output. Increasing the inverter output voltage tended to improve system stability, indicating a weakness in the picoemmeter internal circuitry.

Testing of the failed picoa= meter resulted in replace-ment of four electron tubes (one weak, one with internal leakage and two marginal). Failures of this. type are considered to be within the design limitations of the equipment.

The Technical Specifications and plant design provide for re-moval of one power range flux monitor from service without compromising safety.

3.

Neutron Monitoring Channel No b 12/21/72 - The Log N-Period amplifier in this channel was repaired following variations in the Log N output. The problem was traced to a defective electron tube (internal leakage). Failures of this type are considered to be within the design limitations of the equipment.

The Technical Specifications and plant design do not require this instrument to be in service when reactor power is above 5% of rated power.

h.

Neutron Mcnitoring Channels No 6 and 7 a,

8/3/72 - Failure' of both startup channels occurred dur-ing the reportin6 Period.. Mechanical damage to the chamber / cable compon-ents was the major. cause, requiring considerable maintenance for correction.

' For details please reference AEC letter dated September 13, 1972. Reactor start-up did not begin until after repairs had been effected..

b.-

8/31/72 - The log count rate meter in Channel No 7 was replaced with the' spare unit due to failure of the period amplifier indie3 tion..

w w

19 Eench testing of the failed log ecunt rate =eter resulted in replacement of three electron tubes. Failures o. this type are cen-sidered to be within the design limitatien of the equip =ent.

The Technical Specifications and plant design do not require this instrument to be in service when the reactor is at pcver.

c.

10/5/72 - The proporticnal counter in Channel No 7 was replaced following excessive count rates during shutdovn condition.

This proble= appeared following the shutdovn of Septerter 30.

The comt rate gradually began increasing until it reached a level of approxicately 150 to 200 counts per second (nor:al count rate is 20 to 30 counts per second).

Extensive testing of several ccewnents (leg count rate neter, current pulse amplifi-r and coaxial cabling) was perforced, as oscilloscope 4servation of

'.e log count rate reter input pulses in-dicated additional pulses approximately cne half the a=plitude of the neutron pulses. This was first thought to be loop escillation.

Ecv-ever, after cha ber replacement corrected the prcble=, it is new felt that the additional pulses were extremely high amplitude ga=ra pulses.

This chamber had been inserted to the "in" position pricr to reactor shutdown and was subjected to extremely high neutren and ga m levels which nay have caused tne proble=. Testing of the cha=ber during a prolcnged shutdown remains to be done.

Reactor startup did not begin until after repairs had been effected.

5.

Motor Generator Sets 10/12/72 - The outboard flywheel bearing en M3 set No 2 was replaced due to noise and vibration. Inspecticn of the bearing that was re=oved revealed a rough soot en the outer race.

All equip =ent povered throu6h the No 2 totor generator set was transferred to 1-Y daring the period cf time when the No 2 =otor generator set was out of service.

6.

Ee-etor Protectien syste=

f 10/26/72 - Cleaned contacts and checked operation of rer.ctor protection syste= sensor PS/RE07D (reactor high pressure) following sev-eral spu-ious trip signals frc= same.

20

?

is When a sensor is disconnected for maintenance, automatic tripping of the safety channel to which the sensor is connected will This action provides for reactor safety during the repair.

occur.

B.

Radioactive Effluent Monitoring Systems 1.

Off-Gas System 11/16/72 - The purging air supply solenoid valve used with the off-gas monitor became inoperative and was repaired. This valve is normally in the closed position and its failure or isolation for repair would not affect safety cpe'tation.

2.

Stack Gas Radiation Monitor System 8/3/72 - The differential discriminator in this system a.

was repaired. following erratic operation of the single isotope channel.

Repairs consirsted of electron tube replacement. Failures of this type are considered to be within the design limitation of the equipment.

b.

10/19/72 - The linear amplifier in this system was re-Bench placed with a spare unit following periods of erratic operation.

testing of the failed emplifier esulted in replsicement of several elee-tron tubes (two shorted, one very weak and two marginal). Failures of this. type are considered.to be within the design limitation of the equip-ment.

c.

12/21/72 - The log count rate meter in the gross count channel was repaired following loss 'of sensitivity to a changing count rate. Repairs consisted of electron tube replacement. Failures of this l'

type are considered to be within the design limitation of the equipment.

l Removal of this system from service is permitted by the I

Technical Specifications provided repairs are made promptly and the sys-tem is returned to service. The off-gas monitors provide backup for this monitoring system.

3.

Liould Process Monitoring Systems I

a.' 8/2h/72 - The linear count rate meter in the discharge canal liquid process-monitor was repaired following erratic operation'.

Repairs consisted of ' replacement of several marginal electron tubes.

f. (

LThe unit was returned'to service following a complete calibration check.

Failures of _ this type -are considered to.be within the design limitation of the equipment, i

l L

i P

21 b.

9/7/72 - The linear count rate meter in the discharge canal liquid process monitor was replaced with a spare unit following upscale failure of the unit. Repairs to the failed unit consisted of replacement of a shorted series regulator tube in the negative 150-volt power supply. Failures of this type are considered to be within the design limitation of the quipeent.

c.

9/21/72 - The linear count rate meter in the main conden-sate liquid process monitor channel was replaced following upscale failure.

Bench repair of the failed monitor consisted of replacement oi three electron tubes, two of which were marginal and one exhibiting signs of internal leakage. Failures of this type are considered to te within the design limitation of the equipment.

d.

11/16/72 - Failure of the high-voltage power supply for the sphere cooling water monitor was traced to a defective capacitor in the voltage regulating circuitry. The power supply was repaired and returned to service. Failures of this type are considered to be within the design limitation of the equipment.

Removal of these syste=s froa service are per=itted by the Tech-nical Specifications provided repairs are pro =ptly made and the syste=

is restored to service.

C.

Control Rod Drives and Associated Systers 1..

Control Rod Drive Scram Accu =ulater System

'10/5/72 - A leak between the upper and lover accuculater a.

halves in control rod drive system F-3 vas repaired through installation of a new O-ring and seals.

b.

10/12/72 - A leak between the upper and lover accu =ula-

-tor halves on control rod drive system B-3 was repaired through installa-tion of a new O-ring and seals.

c.

11/16/72 - Spurious accuralator leak alarms were received on the C-2' accumulator leak detection system. The proble= vas corrected through adjustment of the detection system float and switch system.

d.' 12/7/T2 - Investigation of the auxiliary. relay (IS'A type) '

(

'in.E-E rod drive accumulator alarm scheme revealed high contact resis-tance.on the "B" ' contacts of the unit. This results in arcing and radio e

v wr,

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22

(

frequency interference that effects many pulse counting systems in the plant. Plans vill be made to burnish and clean the contacts of all 32 relays to reduce these problems.

The repair of these accumulator leaks was conducted with the reactor at operating pressure. Under these conditions, the primary hydraulic source for drive scra-ing ccces frc= the reactor vessel.

This design feature per=its repair of an accu =ulator without affecting reactor safety.

2.

Control Rod Drive Terrerature 11/20/72 - The thermoccuple circuit for rod drive D L failed en November 2h. Temperature of this d-ive is being manually observed until repairs can be made.

3 Control Rod Drive Seals a.

11/12/72 - The G-rings were replaced on drives E-1 and B-3 due to leakve. The leaking 0-rings were the new silver coated type installed during the last~ refueling outage. For details please refer to the 16th Se=iannual Report of Operations dated August 30, 1972.

They were replaced with the standard teflon coated type which appears to be the most satisfactory type 0-ring at this time, b.

11/23/72 - The sealing 0-rings were replaced on drive

.C-5 because of excessive leakage. This was the last of six drives on which silver coated 0-rings had been installed. All drives nov have the _ standard teflon coated type 0-rings installed on the drive flanges.

The reacter was in cold shutdown while this system was being repaired.

D.

Containment and Associated Isolation Syste=s 1.

Reactor Clean-Up and Fuel Pit ' Drain Isolation Valve a.

7/6/72 - The solenoid control. valve SVh876 on the in-side isolation valve CVh028 was replaced. Cause of the'calfunction was deter =ined to be' a burned out solenoid coil. The failure of this cen-trol valve caused the' isolation valve CVLO28 to close and no further

-safety precautions were_ required.

b.

8/31/72 and 11/23/72.--Additional failures of this sole-l noid valve occurred. 'Ibese failures are discussed in detail in the' Performance Characteristics section of this report.

23 4

~ i 2.

Containment Sphere Heating and Cooling Systen 12/1/72 - The "A" unit heating and cooling heat exchanger tube bundle developed leaks. Inspection revealed two leaking tules and extensive crud buildup in the tube bundles. All tubes were rodded out and the two leaking tubes were plugged. A visual leak test of the tube bundle was perfomed using the heating and cooling systen supply fluid with the tube side bonnet off. The bonnet closure was similarly leak 4

tested using steam from the plant heating boiler.

The unit was isolated during the repairs and no additional precautions were required to provide for reactor safety.

E.

E=ergency Systens 1.

Emergency Diesel Generator 9/14/72 - Failure of the plant emergency diesel gener-a.

ator to come up to rated voltage during a weekly operational test was traced to a shorted exciter armature. The armature was sent out for revind and was. reinstalled by plant caintenance men. On return to ser-vice, a satisfactory load test was perforced at the 200 kW cachine rating with no' descrepancies being noted.

For additional discussion of this failure, see the Performance Characteristics section of this report.

b.

11/12/72 - A new Worthington Model CN FD:-8h cooling water pu=p was installed in ple.ce of the old unit.

Pu= ping capacity and pri=ing time tests were perforced satisfactorily and consisted of 63 gym and 15 seconds, respectively.

The -reactor was in cold' shutdown during the repairs and no additional precautions were required to provide for reactor safety.

F.

Reactor Coolant Pressure Boundary 10 CFR 50.2(v) l

)

~ Reactor Cleanup System Punp

1.

11/5/72 - The clean-up system pump failed and was replaced with the syMen spare (in accordance with Quality Assurance require -

ments). Inspection of the-failed punp revealed the motor vindings to:

be grounded which was caused by failure of the pump bearings'. The failed unit pump casing has been returned to the vendor for a new sta-

- (-

-tor, rotor and bearing assembly. ' This system was valved out of the primary system during replacement of the pump :and a' biannual bearing inspections schedule has been set up to prevent recurrence of this problem.

2-i VI.

CH^."GES, TrSTS AND EXPERI'EiTS Facility Changes Ferformed Pursuant to 10 CFR 50.59(b)

This section contains changes, tests and experiments that do not require Ccc: mission authorization pursuant to 50.59(a).

C-lc9 - Eliminates Unused Valves and Piping in Stack-Gas Sa rling Systen This change rencves certain unused valves and piping fre the stack-gas sarpling system. The safety evaluaticn ccncluded that the re-moval of the extra valves and piping would tend to i= prove the reliability of this syste= and that the chang 2 does not involve a change in the Tech-nical Specifications nor is there an unrevievad safety questien.

C-195 - Re=cves the Voltage Stabilizing Transferner in the Air E.jector Off-Gas Mcnitor Pcver Sources This change removes the volta 6e stabilizing transfer:er in each of the air ejector off-gas systers. The safety analysis concluded that external voltage stabilizatien is not required since the present off-gas instrumentation contains its evn internal voltage stabilizatien equiprent.

Therefore, it was concluded that the change does not involve a change in the l'echnical Specifications nor is there an unreviewed safety question.

C-109 - Adds Test Valvine to the Sphere Vent Valve Centrol Syster This change involves the addition of valving to facilitate leak rate testing of the vent valve sir supply systen and the vent valve. The safety analysis concluded that this addition would i= prove reliability through ease of testing of the systen and that the change does not involve a change in the Technical Specificatiens or an unreviewed safety question.

Test Conducted 11/10/72 - Hot 0 erating Control Rod Drive Scram Times Six centrol rod drives were individually scranmed at reactor pressures varying between 805 psig and 1280 psig to gather additional data concerning the effect of pressure on drive scram times. The scram time test is normally run in the cold shutdown condition as illustrated in Sec-tien D, "Results of Surveillance Test and Inspection Required by Technical Specifications. " The test at various pressures showed that the time re-(

quired for 1007. of blade travel varied between 135 seconds and 1.1.6 seconds.

This test, along with existing data, demonstrates that seras times re=11n within Technical Specifications requirements at any reactor operating pressure.

25 i

The conductance of this test was controlled by written procedros and utilizes the individual withdrawing and scramming of a control blade.

At no time durin6 the test is core reactivity increased. The safety anal-ysis concludes that this test b censistent with the Technical Specifica-tions and does not involve an unreviewed safety question.

Test 7/6/72 6' We Load Rejection Test A 63 We load r?jection test was conducted en Ju.y 6,1972. In an attempt to maintain reactor pressure and flux spikes below the scram set points, the pressure set point of the turbine bypass valve pressure centrollers was modified to step change -37 psi coincidental with 199 ces trip (onset of load rejection). This would insure an i==ediate and con-tinual open signal to the valve. Previously the controllers had been set to open the valve at 10 psi above reactor pressure.

The test results showed that this modification resulted in signifi-cant improvements in the pressure and flux transients associated with such a test. Reactor and drum pressure transients were small (about a 10 psi in-crease) following the 199 OCB trip and stabilized in about 16 seconds at about 24 psi below the initial pressure conditiens. This is close to the calculated change in pressure set point of 37 psi since the original set point was 10 psi above operating pressure. The neutron flux spike was frcn 86 to 101 (17% increase). Extrapolation for a 75 We test leaves 35 margin below the scru set point.

However, the initial cpening of the bypass valve followed by pressure control takeover resulted in a considerable increase of water level in the condenser hot vell. This caused the 4" condensate reject valve to open fully for over two =inutes (1000 gpm) and resulted in loss of suction pressure to the feed-water pu=ps.

This tripped both feed-water pumps and caused subsequent scram on low drum level 85 seconds after the 199 OCB trip occurred.

This operating difficulty is being investigated and modifica-tions will be co=pleted following the results of the investigation.

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4._ c,.,. s,. e g, _,.. e a.3s, e _..,.4.s....<.,, _.>. 4.,.

.e. 3 4. s +.e _..

.3 s.ys%.. " ',.. ' ' _. *. - * - a 0 _*. *.a. k..~.,.

'2..e.',. ^ _.~..

.e e.a.' c -...' '.. _' ~.. S e s

.s.

d.<.e..

< s..

4..

s..,,,, 4 4. as...,.,, ee..,

..a r....c..,...

..~ m..

de.=~ ~r..*,A.

• * ~. +~

~ 7.0, 17, s ',

= ;O '

• .e $.~ ".., +' ~.. v=s use'

~

• ,. 3 ~,:la +.. o..

.e.c. e g,, e.e.

.e.

- -

• s *.e.

. e '. A' a..-. e ','..*.. '.".a...,. " _"

m.

+,. ass.ew.

t.e..,. 44x n.

.. y...-

Ac*ivation gr.s releases ire c =;esed pri arily o' K-13

~te

-a'.e

.~.# *e.' e.a sa. *s m e*.'a.va'. da.,*. nd a...* a.2 ' s.* *..... r. *.*. e.' *. ^. e._' ' '

. r

thly releasu shevn in igendix A.

1

~he six nuclides t.re:

T. -S n, -87, ~68 and Xe-133, -135 and -13S.

i

{

27

(

I Particulate and halogen releases to the atmosphere are measured by counting particulate and charcoal filters weekly. These filters col-lect stack effluent continuously at a rate of 3 cubic feet per minute.

Deter =ination of release rates in this manner assumes radioactivity is continually being deposited uniformly throughout the week on the filters and, hence, a decay correction to the time of analysis is applied, de-pending on the half-life of the nuclide observed.

Tritium releases to the atmosphere are calculated, based upon measure =ents made in the primary coolant and contain=ent air and using identical concentrations for all releases as follows:

g a.

Off-Gas - A flow rate of 10 cfm containing 100% relative h

humidity and 90% radiolytic gas by volume both at primary coolant tri-1 tium to hydrogen ratio to determine tritium releases both in vapor and molecular for=.

The increase in tritium released to the atmosphere over 15rti is primarily due to the increased tritium concentration in the pri-

=ary coolant. This, in turn, is due to the increased recycling of liquid

{

radioactive vaste back into the pri=ary coolant system.

b.

Turbine Sealing Stea - The measured flow rate at 100%

relative humidity and primary coolant tritium to hydrogen ratio.

-c.

Centainment Ventilation - The measured flow rate and nea-sured containment building tritium concentration.

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

ii. - -Licuid Effluents Liquid waste releases totaled 1.09 curie of radioactive mate-rial._ This. release corresponds to 0.88% of technical specifications limits.- Additionally, 10.h curies of tritium were released correspondirs to 0.003% 'of 10 CFR 20'permis'ible concentrations in the discharge canal.

s Liquid Effluent Calculationa Methods The release pathway to Lake Michigan for all liquid effluents

- is_through the plant's condenser circulating water discharge canal. A flow rate of h8,000-52,000'gp= dilution,for liquid effluents is obtained

-thrmgh the use of the condenser circulating water pumps, twoat 24,000

/,p= each, and house service water pu=ps, two at 2,000 gp= each.

Each collected tank of liquid-is sampled, analyzed for radio-active content, and discharged at a contro11ed' rate to assure that y 94

-i%

e y

a.

y y

y

,ec,,4

28 i

permissible concentrations are not exceeded in the canal prior to dilution in Lake Michigan durirs the time of discharge. Each sa=ple is analyzed by gam spectrometry to identify as many of the co=ponent nuclides as pos-sible.

(See Appendix B for results.) Per=issible concentrations in the canal are determined from the folleving:

Ci g,

I

~

MPC i

where Ci is the concentration of the ith isotope in the canal at the given concentration measured in the tank diluted by the known canal flow rate.

Those isotopes not identified by ga==a spectro =etry but measured by a gross beta analysis are presu=ed to be Sr-90 and released on that basis. Periodic sa=ples of the batches are then sent to the radiological environ = ental contractor and analyzed for Sr-90 and Sr-89 From concentra-tions of Sr-90 and Sr-89 found in the batches, the total curies released of these two isotopes is calculated and used in calculating the percent of applicable limit in Appendix B.

The remaining unidentified isotopes are

-6 assigned an MPC of 3 x lo uCi/mlperloCFR20. Tritium released are based on previously established average concentrations in both " clean" and

" dirty" waste tanks.

iii. Solid Wastes A total of 1,139,875 curies of radioactive =aterial was shipped l

off site during the period covered by this report. out of this total, 1,035,000 curies were irradiated cobalt and 104,875 curies were solid vaste.

l.

See Appendix C.

VIII. ENVIRoM' ENTAL MoITITORING

i. Environ = ental Survey I

Environmental levels of radioactivity as found in the vicinit'/

of the plant were composed almost entirely of naturally occurrirg radio-active materials. In the vicinity of the circulating water discharge l-canal. radioactive material of plant origin was found. These materials occurred primarily in aquatic organisms. These. levels of radioactive

. materials, however, were extremely lov and pose no threat to the health

,I and safety!of the public. Further, these-levels of radioactive material found in the resident biological community are consistent with levels found in previous years and show no upward trend.

29 1

s The environmental surveillance program includes continuous sa.pling of air for particulate and halogen activity at seven locations including background sa=ple locations at Traverse City and Boyne City, Michigan, about 50 miles south-southwest and 20 miles southeast of the plant, respectively, to determine increased concentrations, if arce, of radioactivity of plant origin. In addition, film badges placed at each of these locations plus six additional locations on the site property boundary measure direct dose in the environment.

In order to obtain greater sensitivity of measurement, a ec -

parative program of film vs thermoluminescent disometers (TLD) was started in late lgr/1. The program consists of placing a film and TLD side by side at each monitoring station for a one-conth exposure period.

Average monthly doses at the site, inner ring and background stations are co= pared and any difference, at the 95% confidence level, is re-ported using standard "F" and "t" tests. The results of these dosimeter analyses are given in Appendix D.

Ynile all the dosimeters reccrd doses frc= natural occurring sources, the dosineters on site can also be ex-pected to receive doses from not only the plume but direct radiatica frcm the plant and shipments of radioactive material to and from the plants.

The site dosimeters showed, on an average, 2.4 1 0.5 =R/=o above the background statien dosimeters. During the same period of tine, the inner ring of dosimeter stations showed a dose rate above the background station dosimeters on only three occasions.

Air sa=ples gathered continuously and analyzed weekly at the stations shown in Appendix D showed no difference, at the 95% confidence level, in level of radioactivity measured at those stations close to the site and those remote from the site. Both particulate filters and carbon cartridges are used to measure potential concentration of radioactive materials resulting from plant operations. From the known meteorological dispersion conditions, the following maximu= concentrations cr.n be cal-culated:

-1 3

Particulates (Dece=ber)

(1.2 pCi/s) x (0.0027) x (5.0 x 10 s/cc )

1.6 x 10-pCi/c

-1 3

Halogens (February)

(1.2.pC1/s) x (0.009h) x (5.0 x 10 s/cm )

=.5.6 x 10-1' pCi/cm3

~

L 30 1

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

Maximum Calculated Minimus Detectable Normal Background j

Concentration pCi/c=

Activity pCi/cm3 Activity pCi/cm3 Release

-1

. Particulate 1.6 x 10" l x 10" 7 x 10

-l'

-13 Halogen 5.6 x 10 2 x 10 Hence, the negative data obtained in the program was expected.

Also, 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. These results are I

shown in Appendix D.

Additional aquatic ss.mples are taken and analyzed during the su==er growing season and these results are also tabulated in Appendix D.

Based upon the liquid release of 1.09 curies of radioactive ca-terial (less tritium) which results in an annual average concentration in t

i the discharge canal of 1.1 x 10-pCi/ml, the analysis of discharge canal water should indicate an increase of radioactive material in discharge

[

canal water samples since the minimum detectable activity for gross beta

-9 measurements is about 5.x 10 pCi/ml or about-2 times-lover than the av-erage concentration discharged. The results shown plotted in Appendix D l

ind'icate an average of about 8 x'10 pCi/ml for the year.

ii. Environmental Dose Calculations l

' Levels of radioactive materials in environmental. media indicate

~ that public.. intake is well.below 5% of that that could result fro = contin-uous exposure to the concentration values listed in Appendix B, Table II, Part 20.

a.

' Atmospheric Releases In' order to predict potential' radiation doses resulting from f

gaseous releases, environmental transport and uptake factors must be known.

L j

A confirmation of these ce.lculated doses is attempted then by measuring l, _

levels of radioactive' materials in the plant's environmental surveillance

. program. In previous reports for the Eig Rock Point Plant, the average yearly meteorological relationship between release rate and downwind t

l

~. -

31 t

concentration of radionuclides at distances from the plant up to SC miles has been used. Using a typical equilibriu= mixture of noble gas release after a 30-minute decay, a concentration of 1.4 x 10 uci/c=3 delivers in air an exposure rate of 1 millirem per hour using semi-infinite cloud geometry. FromthisandtherelationshipX/Q=3.h5x10-l' s/c=3, the release rate required to produce a dose of 500 millire=s per year at the point of =aximum ground concentration was 2.h curies per second. The licensed technical specification limit on the other hand is 1 curie per second. Furthermore, the licensed technical specification limit for particulates and halogens is (1.2 x 1010) x (MPC) which produces a con-centration at the point of =aximum ground concentration some 2k00 ti=es below =aximum permissible concentration allowed in Title 10 CFR 20 for I-131 using the previous meteorological relationship.

The X/q value of 3.45 x 10 s/c=3

-1 was based on the fact that the wind conveys a plume in Sectors 1 and 2 (see Appendix D) during near neutral conditions 21.25% cf the time per Section 915.11 of the FHSR.

Usingthe=eteorologicaldataintheFHSR,X/Qvaluesforeach each sector, have now been calculated which account for the amount of time the different stability classes exist and the wind is blowing in each sector. At the site. boundary, a plume frc= the plant has not yet reached ground. level. Therefore, any dose received from plant releases at the site boundary would be a shine dose frc= an overhead finite cloud.

Per 'Yeterology and Atomic Energy - 1968," using a diffusion mixture, the release rate required to deliver 500 =1111re=s per year at the site bound-aryis0.66Ci/sinthecriticalsectorwhichisSector4(seeAppendixD).

For an equilibrium mixture, on the other hand, the release rate required to deliver 500 millire=s per year at the site boundary is 1.05 Ci/s.

A co=puter =odel is now used to calculate radiation dose re-sulting from plant releases of noble gases. The integrated populatien dose, out to 50 miles, for lgr72 is shown on the followir4 page. The coeputer model utilizes the following:

X/Qvaluesforthefivesectorsareaveragedoverboth a.

(

stability class and wind frequency.

32 1

b.

Doses are calculated for each of the 22 noble gas radio-nuclides and daughter products based on individual decay energies. Total dose is then the sum ation of the individual nuclide contributions, c.

The 1972 population is esticated from the 1770 Census of j

Population on a township basis corrected by the census-determined State of Michigan growth rate of 1.3% per year and includes transient popula-tion as 1/h residents. The total esti=ated 1772 population resides 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day all year at the same location.

d.

The actual mixture found during the weekly off-gas analysis is used for that week's releases and total re. lease is further corrected by daily measurements of off gas.

e.

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

f.

No credit is taken for the meandering of the plu=e before it reaches the different annuli.

4 The calculated radiation dose at the site boundary resulting from noble gas releases was 8.3 millirers. The integrated dose to the population out to 50 miles was 8.2 =an-F.e s.

Doses fron particulate iodine and tritium releases as shovn in Appendix A were negligible co= pared to that received from noble gases due to the conservative limits in the plant technical specifications and the absence of any significant tilk food chain in the nearby area affected by the plant.

i t.

F I

\\

L

33 i

CAIEULATED RADIATION DoGFS FRCN GASEOUS RELEASE January 1,1972 to December 31, 1972 (man-Rem)

Sector h

5 Tetal i

1 2

3 1 Population 13 74 0

10 97

-2

-2 Population Dose.

2.8 x 10.2 7 5 x 10 0.0 1.8 x lo 0.0 0.12

- 3 Population 261 266-o So 72 649

-2

-2 Population Dose 0.35 0.18 -

0.0 6.7 x 10 9 5 x lo 0.69 3-h Population 555 392 0

47 57 l'.051

-2

-2 o 83 Population Dose 0.52 0.20 0.0 4.8 x lo 5 5 x lo 4 Population 3,300 712 o

102 0

k*114

-2 Population. Dose 2.2 o.29 0.0

.8.^ x lo 0.0-2.5 5-10 i

Population 2,CTTk 24 -

0 527 0

2,625

-3 Population Dose 0.62 4.4 x 10 o,o o,1g o,o o,31

'10-20 Population 8,868 390 737 13,928 323 24,246

-2~

-2 Population Duse 0.66-1.8 x lo "

7.2 x lo 1.4 25x10 2.1 20.

323

. 19,763

' Population

9,523 3,h58 1,895 4,564.

-2

-2

-3 Population Dose 0.2o h.7 x 10 5.5,x 10 0.14 7.8 x 10

- 0.44 30-40 5.'703 k

o 34,161 Population..

22,474.

h,o27.

2,877

-2

-2

-2 9 x 10 0.0 0 31

. Population Dose o.20-2.2 x lo 3 5 x 10

40-50.

5

-2 11,941 ' 2 o.

66,757 Population 40,251

-8,770

.3.'795 2

8 x 10 7 9 x 10-0.0 0 33 Population Dose

0.19 2 3 x 10 o-50L 11,304 35,743

.775' 153,P54-Population :.

87,319 18,113 2 Population Dose 5.o

0.86

~. o.20.

2.1 0.18 8.2

-3

-3

'8.1 x 10 Dose.(Rem) 17.1 x 10 -

4.6 x 10~.3 8.3 x 10 site Boundary.

_qay,-4

I

^!

b.

Liquid Releases In order to predict potential radiation doses resulting from the liquid releases, environseL+.a'. transport and uptake factors must be known. A confirmation of these calculated doses is then at-tempted by measuring leveh of radioactive caterials in the plant's environmental radiation surveillance program.

The nearest municipal drinking water supply intake is located in Charlevoix, Michigan which is generally upstream of the pre-valling current flow in Lake Michigan at this location. However, since current patterns do occur that could, at times, carry the discharged water in the direction of Charlevoix, population dose based upon this flow is calculated in the next section of this report. A conservative dilution factor of 800 is taken frc= the point of discharge to the City of Charlevoix based upon the report, " Big Rock Point Hydrological Sur-vey, Great Lakes Research Division, University of Michigan, Special Report No 9," by John C. Ayers,1951.

In addition, the population dose is calculated to the entire population which receives its drinking water from __J.e Michigan, based en a uniform concentration, resulting from plant releases, through-out Lake Michigan. Also, radiation dose to human populations can occur as a result of plant releases through the consumption of fish caught in Lake Michigan.

Utili?.ing the measured values of radionuclides releasr 1 as shown in Appendix B, the following for=ulr_, and the standard can model, drinking water doses can be calculated as follows:

I Ci I g

E (Limiting Dose Re=/Yr)

D

=

(

ij where:

D,istheindividualdoseinRem/yr, Ci is the average concentration in Lake Michigan of the individual nuclides measured, in pCi/ml, MPC is the concentration of each nuclide measured required to produce the limiting dose at continuous intake in pCi/c1 and

,i Limiting Dose is the dose produced at continuous exposure to MPC concentrations.

35 4

In calculating ingestion dose from the consu=ption of fish, an equation 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 contrast to the 2200 ml of fluid consumed daily by the standard man. This, in effect, alters the MPCi by 50/2200 or 0.0227.

The calculation of individual doses, both from drinking water and consu=ing fish, are per the previous for=ula while integrated pcpula-tion doses in =an-Rem are calculated utilicing the following para =eters:

a.

For drinking water, the individual doses are su==ed over the entire population that receives its drinking water 'ror Lake Micnigan with discharge canal flow appropriately mixed with the lake. This is approximately 10 million people of which approxi=ately 7 tillion reside in the Chicago metrqpolitan area.

b.

The population dose due to d2 inking water to Charlevoix residents is based on a population of 3500 people.

c. *For fish consu=ption, the average concentration in Lake Michigan vater, resulting fro = plant releases, is used with a bicaccurula-tion factor to determine the average concentration in fish.

d.

Fish do not reside cont #.nuously in the discharge canal but migrate. This can be seen in the following table which co= pares the fish consu=ption dose based on the discharge canal water concentration and the appropriate reconcentration factors to the fish consu=ption dose calculated from actual concentrations in fish caught in or near the discharge canal.

Population doses based upon drinking water from the Charlevoix municipal system was 0.0015 can-Ren and tctal Lake Michigan drinkirg water consu=ption population dose was O.47 man-Rec. The consu=ption of all of the Lake Michigan fish harvested. resulted in a population dose of 0.24 can-Re=.

f

• ERG Sp cial Report No 1, " Trace Element Distributions in Water, Sediment, Phytoplankton, Zooplankton and Benthos of Lake Michigan: A Baseline Study With Calculations of Concentration Factors and Buildup of Radio-isotopes in the Food Web," May 1972.

L.'

36 t

l

• Fish Censu=ption Fish Consu=ption Average Con-Average Con-Dose Calculated Dose Eaced en centration in centration in From Discharge Canal Concentraticn in Discharge Canal Sa= pled Fish Concentration Sanpled Fish Tsotcpe (UCi/=1)

(UCi/g) cRe=/Yr rRe:/Yr

-1U I-131 5.0 x 10 9.5

-9 Cs-134 2.8 x 10 8.h Cs-137 2.8 x 10~9 2.h x 10 39 0.14

~7 Co-58 1.7 x 10~

2.0 Co-60 8.3 x 10~

0.31 Zn-65 2.2 x 10~

3.4 x lo-o.023 0.0039

~11 Sr-89 4.5 x 10 0.027

~11 Sr-90 3 0 x 10 1.5

-9 Others 3.1 x 10 5.1 Tne fish consu=ption dose calculated frc= discharge canal concen-trations is 10 to 20 times as large as the fish consu7 tion dose calculated from actual concentrations in fish.

As a censure of total environ = ental i= pact, the radioactive liquid releases from the plant are averaged over the entire lake and then used to determine the population dose fro = fish caught throughout the entire lake and total water consumed from the lake.

Both of the dose calculations are conservative in that:

a.

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

b.

No credit is taken for precipitation and deposit in s~ di-ment or uptake by life for=s other than fish which are seen to occur by l-the data shown in Appendix D.

c.

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

Results are shown in the following tables.

• Utilizing concentration factors found in ERG Special Report No 1, " Trace

(

Element Distributions in Water, Sediment, Phytoplankton, Zooplankton and Benthos of Lake Michigan: A Baseline Study With Calculations of Concentration Factors and Buildup of Radioisotopes in the Food Web,"

May 1772.

7 1

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CONOTNFRS RVFT4 CCHpANY Big Rock Nucienr Ibwer }>1srt Calculated Radiation Doses From Liquid Effluents - Fish Consumption Dose 1/1/72to12/31/72 Avg Concentration Avg Concentration U'

MFC (1)

Critical Bioaccumulation in Lake Michigan in Fish i

i i

Tbpulation Dose Vector Isot w e i

Organ Factor (UC1/ml)

(pC1/r)

(mrem /Yr)

(man-Rem)

Fish Zn-65 k.40E-03 Whole Body 900.0 1.69E-15 1.52E-12 1.73E-07 0.00010 Fish I-131 1 32E-05 Thyroid.

500.0 9 95E-15 4.9dE-12 1.8oE-04 0.11199 Fich Cs-134 3.96E-04 Whole Body 2360.0 4.51E-14 1.06E-10 1 34E-04 0.07775 Fi:h Cs-137 8.80E-04 Whole Body 2360.0 7.26E-ll' 1.71E-10 9.74E-05 0.05784 Fish Co-58 4.40E-05 cl Tract 330.c 1.86E-15 6.15E-13 2.10E-05 0.01245 Fish Co-60 1 32E-03 OI Tract 330.0 1.55E-14 5 10E-12 5.80E-06 0.00344 Fich Sr-89 1.32E-04 Bone 80.0 6.74E-16 5 39E-14 4.08E-07 0.00024 F1:h Sr-90 1 32E-05 Bone 80.0 k 52E-16 3.61E-14 2.29E-05 0.01358 Fith Others 1 32E-05' Whole Body 80.0 1.00E-13 8.03E-12 1.64E-04 0.07753 Total Whole Body 3 96E-04 0.23522 Thyroid 1.89E-04 0.11199 GI Tract 2.68E-05 0.01589 Bone 2.33E-05 0.01383 Mtzimum concentration for fish MIC. (2200/50)

ERG Special Report No 1, " Trace Element Distributions in Water, Sediment, Phytoplankton, Zooplankton and Benthos of Lake Michigan:

A Baseline Study with Calculations of Concentration Factors and Buildup of Radioisotopes in the Food Web."

U2ing 23,873,689 pounds or fish harvested from Lake Michigan in 1970. This number includes both commercial and sports catches as thewn in Appendix D minus alewives which are not generally coneumed.

5 This compares to an average background and medical radiatien desa of 0.215 Pem/yr/ person or 1.3 x 10 man-Rem for the population necessary to consume the !Ake Michigan fish catch at a rate of 50 g/ day / person.

1 39 t

IX. OCCUPATIONAL pERS0KNEL RADIATION EXPOSURE I.

Film Badge Recults for No=al Plant Personnel mBem Dose 7/2/72 - 7/30/72 7/31/72 - 8/27/72 8/28/72 - 9/24/72 0 - 100 Maint. 4-Oper. 13 Maint. 2-Oper.

9 Maint. 7-oper. 17 Supv. 15-Tech. 4 Supv. 16-Tec h.

5 Supv. 17-Tech. 8 101-500 Maint. 5-Oper. 7 Maint. 6-Oper.11 Maint. 0-Oper.

O Supv.

2-Tech.

6 Supv.

1-Tech.

5 Supv. 0-Tec h.

O 501-1250 Maint. 0-Oper. O Maint. 2-Oper. O Maint. 3-Oper.

3 Supv. 0-Tech. O Supv. 0-Tech. O Supv. 0-Tech.

2 1251-2500 Maint. 1-Oper. O Maint. 0-Oper. O Maint. 2-Oper. O Supv.

0-Tech.

O Supv.

0-Tech.

O Supv.

0-Tech. O 2501-5000 Maint. 0-Oper. O Maint. 0-Oper. O Maint. 0-Oper. O Supv. 0-Tech. O Supv. 0-Tech.

O Supv. 0-Tec h.

O mrem Dose 9/25/72 - 10/29/72 10/30/72 - 11/26/72 11/27/72 - 12/31/72 0 - 100 Maint. 3-Oper. 12 Maint. 5-Oper.13 Maint. 3 -Oper. 9 Supv. 18-Tech. 7 Supv. 14-Tech. 4 Supv. 16 -Tech. 5 101-500

.Maint. 8-Oper. 7 Maint. 3-Oper. 6 Maint. 9 -Oper.10 Supv.

0-Tech.

3 Supv. 4-Tech. 6 Supv. 2. Tech. 5 501-1250 Maint. 1-Oper. O Maint. 4-Oper. O Maint. O -Oper. O Supv.'

0-Tech.

'O Supv. 0-Tech.

O Supv.

O -Tech. O 1251-2500 Maint. 0-Oper. O Maint. 0-Oper. O Maint. O -Oper. O Supv. 0-Tech. O Supv. 0-Tech. O Supv.

O -Tech. O 2501-5000 Maint. 0-Oper. O Maint. 0-Oper. O Mtint. 0 -Oper. O Supv. 0-Tech. O Supv. 0-Tech. O Supv.

O -Tech. O

O e

/

X.

RADI0 ACTIVE LEVELS IN PRII;CIPAL FLUID SYSTEMS Minimum Average _

Maximum A.

Primary Coolant Reactor Water Filtrate ("}

-1

-1 pCi/ml.

1.16 x 10 2 90 x 10 2 90 Reactor Water Crud (C) pCi/ml/ Turbidity Unit 5 80 x 10 174 x 10-1 2 90 x 10-1

-2 Iodine Activity

-1 pCi/ml 3 x 10-2 6 x 10-2 2 x 10 B.

Reactor Cooling Water System Reactor Cooling Water (*}

pCi/mi1 2.18 x 10-2 2 90 x 10-2 4 36 x 10-2 C.

Spent Fuel Pool Fuel Storage Pool ("

-3 1.02 x 10-2 pCi/ml 1.02 x lo 4 36 x 10 Fuel Pool Iodine pCi/mi 15 x 10-7 3 x 10-5 4 x 10-3

("}A counter efficiency based on a decay scheme consisting of one gamma photon per disintegration at o.662 MeV used to convert count rate to microcuries.

All count rates were taken at two hours after sampling.

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

(#) Based on APHA turbidity units and 500 ml of filtered sample.

CONSUlFJS POWER CoMPAh7 By kk

&2 Nuclehr Licensing Adminis{rator Date: February 28, 1973 Sworn and subscribed to before me this 28th day of February 1973 l

f& { kwad Notary Public, Jackson County, Michigan My ecmunission expires June 20, lgr(6 N

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00R ORIGINAL

s-Appendix C; Transfer of Radioactive Material Shipment Transfer No.

Date From Transfer To Radioactive Material 249'

.1/14/72.

DPR-o G.E., San Jose, Calif.

3 In-Core Chambers 0 3 mci SdM-54 250 1/28/72 DPR-6

NRL,-Washington, D.C.

Irradiated Versel Specimens 109.6 Ci 8-1393-2 A-66 251 1/26/72' DPR-6 Battelle Columbus, Ohio 3 Irradiated Fuel' Rode 3324 Ci SNM-7

.Y 252 2/1/ 72 DPR-6 Battelle Columbus, Ohio 1 Irradiated Fuel Rod 1108 Ci SNM-7 253 2/4/72 DPR-6 Isotopes Inc., New Jersey -

Reactor Water and 0.2 mci 29-55-6 condensate samples 254 2/7/72

. DPR-6 '

-G.E. Val. 0017-60 Feed-Water Crud and 0.1 mci i

California.

Filtrate

-255 3/16/72 DPR-6 G.E. North Carolina 3 00 Fuel Rode 40.1 mci 2

SNM-1097 256 3/30/72 DPR-6 NECO. 36-NSF (A-11) 5 Heat Exchan'ers 1.25 Ci g

.Morehead, Ky.

257 4/6/72

_DPR-6 NPI, 19-12667-01 Irradiated Cobalt ~

542,000 Ci 258

-4/J8/72 DPR-6 NPI, 19-12667-01 Irradiated Cobalt 493,000 Ci 5

1

s

'l 4'

Shipment;

. Transfer.

No.

Date From' Transfer To Radioactive Material- -

259~

t7-10-72 DPR-6 G.E.-San Jose, Cal. SNM-960 8 Irradiated Fuel Rods 29,400 C1 260 7-10-72'

-DPR-6L

G.E. San Jose,. Cal. SNM-960

' Liner for T-2. Cask 2 mci 3

261; l7-11-72

- DPR-6.-

NECo,' Ky.-16-NSF-1

-100 ft resins 95 Ci 3

'262' 13-72

'DPR 6 NECo, ' Ky. 16-NSF-1 50 ft resins 62 5 Ci 263~

T-13-72

- DPR 6

.HECo,.Ky. 16-NSF-1 50.55-gal barrels 92 mci 3

26hE J-7-13-72

-DPR-6

NECo, : Ky. 16-NSF-1 100 ft resins les Ci 265

'7-17-72 DPR 6

.G.E.-Vallecitos, Cal, h irradiated fuel rods 24,700 Ci

-SNM-960 3

.: 266 -

T-17-72

..DPR 6.

NECo, Ky..16-NSF-1 100 ft resins 100 Ci 3

^

-- 267 '

7-21-72 DPR 6-'

NECO, Ky. 16 NSF-1 100 ft resins 80 C1 268-

.7-27-72 DPR 6 G.E.-Vallecitos, Cal.

4 control mds 24,700 Ci SNM-960 269 7-27-72

.-DPR-6 NECo, Ky.16-NSF-1 4 contml rod blades 2h00 Ci 3

270 8 8-72

' DPR 6 NECo,'Ky. 16-NSF-1 100 ft resins 100 Ci 271. 9-72 DPR-6

.NECo,;Ky. 16-NSF-1 h contml rod blades 2h00 Ci

?272.

8-lh-72 DPR-6 NRL-Wash. D.C. 81393-2 Vessel coupons 133 5 Ci 3

273.

"8-15 DrR-6 NECo, Ky. 16-NSF-1 100 ft resins 100 Ci 27h' 8-18-72

.DPR-6 NECo, Ky. 16-NSF-1 7 pieces fuel channel h700 Ci e

275 8-18-72 DPR 6 NECo, Ky. 16-NSF-1 47 55-gal barrels 77 4 mci r

[ 'j b

S' hipment.

~ Transfer No.

Date From

-Transfer To Radioactive Material 3

276

.8-18-72

- DPR 6 NECo,'Ky. 16-NSF-1.

50 ft. resins 50 ci 3

277 8-18-72 DPR 6 NECo, Ky.-16-NSF-1 50 ft resins 60 ci 3

278 8-18-72

. DPR 6

.NECo, Ky. 16-USF-1 100 ft resins 100 ci 3

.279 8-21-72 DPR 6

'NECo, Ky. 16-NSF 50 ft resins 32 ci 3

280~

8-22-72

. DPR-6

-NECo, Ky. 16-NSF-1 50 ft resins-32 ci 281

'8-22 DPR 6 ND:o,LKy. 16-NSF 9 55-gal barrels 47 5 moi 282 8-23-72'

.DPR 6

.NECo, Ky. 16-NSF-1 100 ft3l resins 64 ci 3

283

.8-24-72 DPR 6 NECo, Ky. 16-NSF-1 50 ft resins 32 ci 3

284

.8-25-72

DPR 6.

NECo, Ky. 16-NSF-1.

50 ft resins 20 ci 3

-285; 25-72 DPR 6 NECo, Ky.'16-NSF-1 100 ft resins 64 ci 286 8-31-72 DPR 6

_NECo,.Ky. 16-NSF-1 5 pieces fuel channels; 1620 C1 5 in-cores

-287'

_9-25 DPR 6 NECo, Ky. 16-NSF-1 5 pieces fuel channels; 1164 Ci Misc 288'

.' 9-27-72 '

-DPR 6 NECO, Ky. 16-NSF-1 3 control rod blades; 1h63 ci 8 B C rods; misc g

289 9-29-72

.DPR-6 NECo, Ky. 16-NSF-1 7 pieces fuel channels; 283 ci 3 orifices l'

290 10 6-72

. DPR-6 NECo, Ky. 16-NSF-1 7 fuel channel pieces:

8 orifices; 6 stiffeners; 12.8 ci u

291 10-11-72 DPR 6 NECo, Ky. 16-NSF-1 2 control rod hlades; 2h00 Ci l

5 support tubes l

t

Shipment ~

Transfer.

No..

Date From Transfer To Radioactive Material 292: 12-72

.DPR 6L

'NECo, Ky. 16-NSF-1 7 fuel channel pieces; 9 orifices; 8 stiffeners 34T Ci 293 l10-26-72

'DPR 6 NECo,'Ky. 16-NSF-1 3 control rod blades; 2 fuel channel pieces 3014 C1 294

~10-26-72L DPR 6 G.E.,-Vallecitos, Cal.

2 orifices, 2 transition 260 Ci

-0017 60 pieces; 295 Jil-3-72

--DPR-6 NECo, Ky. 16 NSF-1 7 fuel chtinnel pieces; 47 Ci 13 orifices; 4 stiffeners; 5 plugs 1

296

~ 11-8-72:

DPR 6' NECo, Ky. 16-NSF-1 7 fuel channel pieces; 2 Ci

[

3 oririces; 7 channel plugs

<297 11-15-72

-DPR 6 NECo, Ky._16-NSF-1 7 fuel channel pieces; 13 Ci 7 channel plugs

[

298 11-20-72

-DPR-6 NECo, Ky. 16-NSF-1 7 fuel channel pieces; 1.6 Ci l

7. channel plugs 299 11-27-72 DPR-6

.NECo, Ky. 16-NSF-1 7' fuel channel pieces; h Ci r

300 11-29-72

.DPR-6 NECo, Ky. 16-NSF-1 7 fuel channel pieces; 0 7 C1 301.

.12 6-7?

DPR-6 NECo, Ky.'16 NSF-1 7 fuel channel pieces; 24 3 Ci 1302

12-11-72

.DPR 6~

NECo, Ky. 16 NSF-1 7 fuel channel pieces; h8 9 Ci 303-12-15-72 DPR 6 NECo,'Ky. 16 NSF-1 7 fuel channel pieces; 60.3 Ci 304 12-18-72

.DPR 6 NECo, Ky. 16-NSF-1 7 fuel channel pieces; 65 8 C1

-305 12-28-72:

DPR 6 NECo, Ky. 16-NSF-1 h transition pieces; h7 2 Ci 3 channel pieces 306.

12_-28 DPR 6 NECo, Ky. 16-NSF-1 110 ft3 high-level waste; 10 5 Ci t

47

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.'. 3:. : qt 6'

.s

• n l

\\.,,.. /l 4 L, l

5

.s i

/,h-g P.

.p g

7 j.

1.,.

.e i.

t.

a i

g

. {,3

} ;i g' -

't "

,.[

i i

rJ g,

~

-m

504, 1

P00R ORIGINAL.

m APPENDIX D (Contd)

LANE M/CH!CA N BIG ROCK SCREENHOUSE:

j

@ LEAD CAVE TSIDE LEAD CAVE-J

_r._

(G)lNFORMATION l I /

'(REACTOR STACKh i SUBSTATION 8[ (

1

/

~$

1 l

8 p-

{

s, g

w h

4 5

o

=

I' m

s o

@,.. / ~ ~~

l C

8 LEGEND y/

s g

bl 0

BIG ROCK POINT 3

SITE MAP O

e' WITH ENVIRONMENTAL MONITORING STATIONS g

I h - LET7ERED S TATION S - F I LM MONITOR ONLY r

mm.m.

8 gp g

L/_ _.._

e -"u"oeac o s'a> >o"s -r i' "o ^'a "o"oa

\\

I i

49 i

BIG ROCK POINT NUCLEAR PLANT-DOSE ISOPLETHS 1

5 C. 0/

d IRECTION NO. 3

/

OFF-SHORE

,/

gan (All land trojectory) sPRsNo

/

'WN U /avsy5,,

f olRECTION NO.4 sto

/

• ft SPRMS / PEN [. PEW TO't"Jt0

~~~h~~~~

[" /gg' KEY

,y5waty, trol y W

~j~~-

DIRECTION NO. 5 ALONG SHORE

( imed trojectory

/

DIRECTION NO. I residue)

. /

ON SHOGE

/

L; (All water trajectory)

CpAR

./

)

l p'/

/

/

4 r

o 3

to M

I P00R ORIGINAL i

~

l

BIG ROCK POI 5T NUCL5AR PLANT-DOSE ISOPLETHS 50 Mi r~r e ~~

a f>;y-6N

- g-

/

/

~

/

4 SCALE O

O MILES P00R ORIGlHAL T

APPENDIX D (Contd)

Sampling and Ariclysio Summary Number of Camples Frequency of Medium Leccription loention Collected Type of Analynia _

Analysis Air Continuous at All 3hh arons Beta, 131 I ueekly Approximately 1 Cih

_ I/tke Water 1 Gal Grab GT 2h Cross Beta, Gross Gamma Monthly 90 13hCs, 137Cs, NMn, 4' #t"#1Y 37, 9

Co, Co,

'Zn,

Fe Well Water 1 Gal Grab ST 12 Gross Beta Monthly Gamma Done Continuouc All h')1 Film Dose i

Monthly 20T TID lbec Fish Grab St, NM, 13 Gru:. Beta, Groon Gamma Semi-Annual Mt McBauha Spectrum Aquatie Biota Grab St, NM, 30 Gruno Beta, Grono Gumma cemi-Annuni Mt McGauha 3pectrum

, n.

\\

4 APPENDIX D (Contd)

'High,-Low and Average. Concentrations' lFor 'H1 hest Average Sampling location 6

~

Location High Low Average Type Type ~of Analysis Air-

-! Gross Beta-Gamma ST

-0 51 pCi/l

0.01 pCi/l O.11

~

~

I-131.

,BC 0.27

<0.2 0.20 Lake Water Gross Beta 3R ST LWO' 18 4.5' 93 Gross Gamma BR ST LWO.

35

<6 11.6

Sr BR ST LWO' 2.4 1.8 2.1

.Cs-137 BR ST.LWO 6

<2 4

Well Water -

Gross Beta-BR ST WW 54 05 69

. Film Dose ST' 19 Millirad o Millirad 12 Millirad

• TLD Dose E

7 5 mR

-0.2 3 6 mR

• In excess - of control ' dosimeter L_-

53 I

Difference in Average TLD Readings mR/ Month Site Vs Background Inner Ring Vs Dackground Month Stations Stations January

+2 93 1 1.45

+0.5 t o.43 February

+7.73 t 3.66

+1.8 t 1.74 1

-March N.D.

N.D.

. 3 95 t 1.50 N.D.

' April ~

+

May

+2.93 1.51 N.D.

' June

+2.87 I 2.85 N.D.

July

+5.15 1.55 N.D.

August TLD Readings Lost September N.D.

+0 93 t o.68 october ~

N.D.

N.D.

November N.D.

N.D.

December..

+2 98 +- 2.73 N.D.

Average 2.37 t' o.51 0.27 1 0.16 1 N.D.:

No' Difference at the 95% confidence level.

TLD Readings : lost by Contractor.

dnu o

C

~

o E

rg D

dknc ua ob rgs V

k v O

cag N

bn i

sr v r ee T

t n U

in O

si ex h

S G

U A

sgn i

d L

6 a

"U e

R J

D LT eh N

gt O

an ro J

eM v/

ARr n

i Y

o e

M cn ere f

R f

i o

D A

R 8

E 0

,c.

F

~

N o

A J

f

/

0 9

7 b

g 3

2

/

O

/

/

TCCm Cy M E P

33 3

l :

I:

s APPHIDIX D (Contd)

Big Rock Point Plant a Site Station Monthly Average Radioactivity e Average of Stations 2-5 Concentration in Air

+ Background Stations 6 and 7 Average

.a IA0

-~

4 90 I

.so 3oo e 7o I,

y 80

'206 B j

p-2 1 to'

.40 I w 54, 8 O

.ao Q

.ns

^

u

\\

7 S.

20 g,

/

S fj y '

'5 g

h j[f/

k

[

wa

fo g

i.

s 0,

w w

u s

s\\\\

a

't42

//

,o,

c:a y

.0 5 g

.04 f

O

.03 y

J g

n

\\\\\\

///

o-

\\J1 p--

.08 JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DEC

g p3 '~ g w.

3m[i3l" 0.

3

@ t C s 4

0 s 3

1 t

4 t e el lt C

nu

[

^/

E IO D

rr ee p

tt u

aa A/

WW

//

/

V 3

N ee O

kk N

aa y

l1 J_

A*

r

\\ x C

O N \\

P

\\

E S

\\

N s

A G

U A

,/V

/

y

/

L t

i U

vr J

tie

) n't t

\\

daca tl aW g,

/

nP o o

in Ctdi N

,Mq (na v

U iRn J

Do o

Pei X

6t Ik a a

/

Dcrr I

oet BRvn y

A w

e A

Wg Ae c

M Ai y n Bl o hC t

n m W o

R M

P A

&,x N

R I

A M

/

/

r

/

B E

F

/ /-/

/

/ f d

N v

A J

000 0 O O 0

0 s 0

o,81 s

s.

i 098 7 G S 4

3 a 2

l 1

pD:ewC3-@~>W

tl 3

gh1 5 n n % gc0* =C 1

.m

.m u

Appendix D (Contd)

Big Rock Samples May 1972 Nuclide (pCi/g)

Camma Beta Zr 65 60 40 54 137Cs 95 134Cs,58co Zn Co g

Mn epm /g*

pCi/g Sample Shore. Minnow Distharge:

0.3 0.3 0.7 0.20 3.1 2 0.3 1/4 Mi.

E.

of'Disch.

0.3~

0.6 1.0 0.17 0.27 3.4 t 0.3 1/4 Mi, W..of Disch.

0.2 0.4 1.0 0.15 3.4 1 0.3 Nine Mile Point (a) 0.1 0.3 1.0 0.11 2.8 1 0.3 1.8 0.06 1.9 1 0.2 0

Mt. McSauba(b) 0.2

. Al'ewi f e 0.1 0.1 0.8 0.09 2.6 1 0.3 Lake Trout 0.5 1.0 0.13 3.1 2 0.3 CError is 1 0.01 or 10%, whichever is greater (c) 3 miles East of discharge (b) 3 miles West of discharge

~E3

.~s Appendix D (Contd)

-BIC ROCK SA!!PLES May 1972 Nuclide (pci/g)

Gamma Beta Sample

137c, 95 134Cs.58co 65 n 60 Zr Z

Co 40g 54 tin epm /g*

pCi/g Crayfish Discharge 0.4 0.1 0.3 0.36 2.2 1 0.2 Nine Mile Point (s) 0.2 0.3

?

0.20 1.6 1 0.2 Mt. McSauba(b) 0.3

?

0.13 4.0 1 0.4 Poriphyton Discharge 1.0 0.8 7

0.9 7.7 104 1 10 1/4 Mi.

E.

of Disch.

1.2 0.3 2.2 61 1 6 1/4 Mi.

W.

of Disch.

0.4 0.2

?

0.67 32 1 3 Nine Mile Point 0.4 0.9 0.45 25 1 2

'Mt.-McSauba 0.2 0.2 1.3 0.44 22 2

Filament Algae Discharge 0.18 51 1 5 1/4 Mi.

W.

of Disch.

0.17 0.2 0.13 19 ! 2 Nine Mile Point 0.1 0.1?

1.2 0.19 13 1 1**

.Mt.

McSauba 0.3 0.8 0.17 17 2

Battom Sediment Disch. Midway 0.5 0.2 0.06 0.4 2.4 9.2 0.9 Disch. Off Shore End 0.5 0.2 0.05 0.4 2.0 16 1 2 Disch. Shore Line 0.4 0.2 0.05 04 1.5 12 1 1 CError is i 0.01 or 10%, whichever is greater t

a* Abnormal amount of solids (c) 3 miles East of discharge U>

.(b) 3 miles West of discharge

(?)

Present, but not statistically cignificant.

4

? Appendix D (Cont'd).

Big Rock Point: Samples

November 1972 Nuclide(pCi/g)

-3,,p1, cs--

95

. 58 65 60 h0 Sh Q1 137 Zr co Zn co K

gn c

' Shore Minnows 0 32 1.0 0.22!O.02 2.1 0.2

' Discharge:.

0.2h 1/hMi! East 0.82

-O.33-

,0 98 2.0 0 5810.06 1 3*0.2 l

'1/hMiWest 0.0910.01 0.810.2

.Niise Mi Ptz 0 7610.08 0.810.2

.Mt McSauba'

_0.221 0.26.

1.2 0.1710.02 0 710.2 l

Crayfish Discharge _

0.44.

{0.11 0 52 0.48 0 3610.04 2.610 3 1/hMi' East ?O.h3 0.48 0 3210.03 2 3iO 2 1/hMiWest E0 34-0 38 0 98 0.2410.02 1.6!0.2 Nine Mi.Pt

.0.19 0.85 0.1510.02 2.610 3 L

Mt'McSauba-

'O.16 0 74

( 0.1 1.h10.2

'Periphyton Discharge 0.41 0.19-0.16 0.h2 0.17t0.02 18 2

- 1/h Mi East

_0 56 0.19 0.21 0.21 1 310.01 1011 1/4MiWest' 0 55.

0.11 0.18 0 33 0.6920.07 1211

.Nine Mi Ptl

'O.22 0.07 0.06 13 0 3410.03 9il Mt McSauba 0 341

'O.06 0.05 1.1 0 3810.0h 11 1 Filament-l Algae Discharge 0.24 0.09 0.11 0.27 0.03 6.1!0.6 I1/h.Mi East. 0.20 0.13 0.17 0.hh 0.h520.05 3 8f0.4 8

1/h Mi-West 0.15 0 93 0.21 0.02 2.hto.5 Nine Mi Pt-

.O.34 1.1 0.hhio.01 3 810.h Mt McSauba 0.07 0.26 0.70 0.18f0.02 h.7do.5

60 APPENI/IX D (Contd)

• Michigan Department of Natural Resources Fisheries Division Lansing, Michigan Sports Catch, Lake Michigan and Anadromous Streams, 1970 Species Number Caught Estimated Total Weight (1bs)

Perch 1,700,000 283,333 Walleye 69,000 207,000 Bass 246,000 492,000 Panfish 1,300,000 260,000 Northern Pike lh6,000 292,000 Suckers 482,000 1,446,000 Smelt 2,800,000 280,000 Lake Trout-2k5,000 1,715,000 Rainbow Trout 285,000 1,425,000 Brown Trout 168,000 840,000 Brock Trout 125,000 250,000 Coho Salmen 534,000 5,340,000 Chinook Salmen 180,000 2,700,000 Other Species 368,000 368,000 Total -

8,648,000 15,898,333

• Unpublished 1970 data from postcard census program of the Michigan Department of Natural Resources, Fisheries Division.

+

61 i

APPENDIX D (Contd)

• Michigan Department of Natural Resources

(

Fisheries Division Lansing, Michigan CG00ERCIAL CATCH, IAIE MICHIGAN - - 1970 Estimated Total Species Weight (1bs)

Alewives 5,981,415 Bullheads 610 Burbet 51,261 Ca?.p 2,394 Chubs 4,028,340 Herring 676 Lake Trout 89,939 Menominee 161,987 Perch 22 Pike 65 Rock Bass 35 Sauger 1

Sheepshead 12 Smelt 1,700,365 Suckers 521,807 Walleyes 7

White Bass 1

Whitefish 1,417,834 13,956,771 l

Taken frue: GREAT IAKES FISHERIES 1970 data taken from December issue of MicM gan, Ohio and Wisconsin Landings.

_