ML20030A500
| ML20030A500 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/30/1972 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| References | |
| NUDOCS 8101090776 | |
| Download: ML20030A500 (40) | |
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D N Ml Docket No 50-155 A
co Report of Operation of Big Rock Point Nuclear Plant License No DPR-6 January 1, 1972 Through June 30, 1972 I.
SIM4ARY OF OPERATIONS A.
Power Operation The Big Rock Point Plant was operating at 43 W e (gross) as of January 1, 1972. The off-gas release rate was averaging approxi-mately 27,800 Ci/s.
p On January 4, a coastdown mode of operation was started in anticipation of the beginning of the refueling outage scheduled for March 17, 1972. A fixed control rod pattern was maintained with 20 notches remaining on rods partially inserted in the core. These rods were left in the core to minimize off-gas and assist in preserving fuel integrity.
The plant operated at an average coastdown rate.. approxi-mately 1.1 MWe per week until January 25 when the turbine tripped on overspeed and the reactor scrammed. The turbine trip resulted when a line fault, which had occurred on the system, was not cleared by the Big Rock Point relaying scheme. As a result, the p.lant became momentarily isolated from the rest of the system and, with essentially no load on the turbine generator, the unit tripped off on overspeed.
The redundant station power supply was also lost for a period of time due to a combination of unusual weather conditions and equiptnent fail-ures. Also occurring at this time was an approximate 2-foot drop ia spent fuel pool-water level and the failure of emergency condenser out-let Valve MO-7063 to close.
(Refer to our March 3, 1972 letter to the AEC.)
The plant was returned to service at hO MWe (gross) on January 28, 1972 afteran80-1/2-houroutage. Off-gas release rate levelsaveragedapproximately27,000pCi/s.
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I The plant continued to operate in a coastdown mode of approxi-mately 1.0 We/veek until February 11 when the unit was removed on a scheduled outage to perform work on the turbine initial-prersure regulator (IFR ).
The 7FR would not regulate the turbine control valves effectively at low-power levels. Minor linkage adjustments improved the IFR response only slightly and prompted the plant personnel to schedule complete in-spection of the unit during the refueling outage.
Following this 8-hour outage, the plant was returned to oper-i
- r. tion at approximately 38.5 We (gross) on February 12, 1972.. The plant operated continuously in a coastdown mode until March 18, 1972 when it was shut down for a acheduled 4-week outage. Electric generation at the end of the period was approximately 34 We (gross). The off gas was averaging 28,100pC1/s.
During the refueling outage, all 84 fuel bundles were dry sipped. Thirty-one were found to be failed. Of these failures, 16 bundles were 4-cycle Type E fuel with average exposure of 10,900 E d/T; 9 bundles were 3-cycle Type EG fuel with average exposure of 12,900 l.
Wd/T; 5 bundles were 2-cycle Type EG fuel with average exposure of 9,200 wd/T; and the remaining bundle was Type C fuel with an exposure well beyond the warranted exposure.
(The E-type fuel has a design life of15,000MF/T.) All 24 bundles of Type'F fuel,-with an average ex-posure of 4,500 E d/T (completed first cycle), were.. sipped with no failures indicated. The fuel crud levels that Big Rock Point has ex-perienced appear to be about the same as the last cycle.
The core was reloaded with 30 new General Electric (GE)
Type F fuel bundles, ' two new J-2 (Jersey Nucle'ar) mixed oxide demonstra-tion fuel bundles and the balance'with the remaining nonfailed GE Type EG, F, EEI-Pu and J-1 (Jersey Nuclear) fuel bundles.
The clean-up system heat exchangers (four regenerative and one nonregenerative) with Cufenloy tube bundles were replaced with new heat exchangers utilizing stainless steel tube bundles. It is felt this will help to eliminate most of the crud deposits on the fuel cladding and thus decrease the rateiof premature cladding failures, fT h
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3 During a test operation of an explosive squib valve on the liquid sodium pentaborate poison syste=, a valve failed to functicn properly.
(Refer to our letters to the AEC dated May 11, 1972 and July 20, 1972.)
Investigation of the control blades with missing rollers indi-cates that the roller wear observed is related to the reactor primary coolant inlet diffuser geometry.
(Refer to our April 25,1972.1etter to the AEC.) Forty-two (h2) new fuel support-tube-and-channel asse=blies were installed in the reactor. The new channels are necessary because of the bowing-in or " oil-canning" effect (caused after several years of operation) which hinders removal and insertion of fuel bundles.
The support-tube-and-channel assemblies airo incorporate flow design modifications which will allow a more even distribution of coolant flow in the lower sections of fuel rods.
(Refer to Proposed Technical Specification Change no 28' submitted on January 21, 1972 to the AEC.)
Due to weld failures at the lower end caps on the outer encap-sulation of'both neutron sources, new zirconium sheaths were fabricated and reincapaulated over'the neutron sources.
(Refer to our April 19, 4
1972 letter to the ?.EC.)
The plant was returned to service at 1703 hoe-s May 13, 1972.
1 lOperationwascontinuousuntilapproximately1500hoarsonMay15,at which time the unit was forced out of service because of excessive leak-age of prisAry coolant being experienced at the B-5 control rod drive flange connection to the reactor vessel.- This drive had been worked on-during the outage-and investigatien revealed that use of a new type silver plated inconel 0-ring was the probable cause of leakage. The f
-silver _ plated 0-ring requires greater compression than does the Teflon-F coated 0-rings for equivalent sealing..
The unit was returned to service at approxicately 0300 hours0.00347 days <br />0.0833 hours <br />4.960317e-4 weeks <br />1.1415e-4 months <br /> on
.May 18, 1972 at a gross output of.53 MWe. Plant power level was reduced a number of. times on May 18 and 19 to permit entry into the recirculating--
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- pump room in an attempt to isolate a leak into the-component cooling
- water system. The, leak was traced to th'e No.1 reactor _ recirculating
- water pump 3/4-inch seal cooling water heat: exchanger. The pump was
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4 shut down and isolated at 0730 on May 19, limiting gross output to 51 We. Plant off gas was averaging approricately 700 pCi/s.
The plant continued to operate at 51 We (gross) until 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on June.10, when the unit was removed from service to replace the Just 3/h-inch heat exchanger on the No 1 reactor recirculating pu._p.
prior to the shutdown, off-gas isolation Valve CV-h015 was closed to determine its isolation capabilities. Clesure of the valve failed to throttle off-gas flow enough to reduce condenser vacuum to the trip point.
(Refer to our June 26, 1972 letter to the AEC. )
The unit was returned to service at 0h00 hours on June 11.
However, the unit was again limited to one pu.cp operation (51 We (gross))
because of failure to establish No 1 recirculation pu=p seal ler.k off flow rate. The plant was shut down at approximately 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> on June 17 to replace the seal cartridge on the No 1 reactor recirculation On June 18, after reaching a power level of 5h We (gross), the purp.
plant was again shut down to replace the leaking off-gas rupture dia-phragm which caused higher-than-normal radioactive airborne conditions in the turbine building.
The plant resumed service at 12hl hours on June 18 and has since operated continuously at a gross output of 63 We (gross).
The off-gas activity was averaging approxi=ately 1,000 pCi/s at the end of this report period.
II.
EFFLUE:TP AND EWIROMETAL MONITORING A.
Introduction and Conclusions 1.
Introduction Releases of radioactive material both to the at=osphere and Lake Michigan from January 1 to June 30, 1972 were well within the facility-licensed limits and the Co= mission's regulations, particularly Title 10, Code of Federal Regulations, Part 20. Environmental levels of radioactiv-ity as found in the vicinity of the plant were co= posed almost entirely of naturally occurring radioactive =aterials. Only in the vicinity of the circulating water discharge canal was radioactive material of plant origin found. These materials occurred primarily in aquatic organirms. These levels of radioactive materials, however, were extremely low and pose no i
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t shut down and isolated at 0730 on May 19, limiting gross output to 51 We. Plant off gas was averaging approximately 700 pCi/s.
The plant continued to operate at 51 We (gross) until 0100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> on June 10, when the unit was removed from service to replace the Just 3/4-inch heat exchanger on the No 1 reactor recirculating pump.
prior to the shutdown, off-gas isolation Valve CV-4015 was closed to i
determine its isolation capabilities. Closure of the valve failed to throttle off-gas flow enough to reduce condenser vacuum to the trip point.
(Refer to our June 26, 1972 letter to the AEC.)
The unit was returned to service at 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br /> nn June 11.
However, the unit was aga'in limited to one pump operation (51 We (gross))
because of failure to establish No 1 recirculation pump seal leak off flow rate. The plant was shut down at approximately 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> on 1
June 17 to replace the seal cartridge on the No 1 reactor recirculation On June 18, after reaching a power level of 54 W e (gross), the pump.
plant was again. shut down to replace the leaking off-gas rupture dia-t phragm which caused higher-than-normal radioactive airborne conditions in the turbine building.
The plant resumed ~ service at 1241 hours0.0144 days <br />0.345 hours <br />0.00205 weeks <br />4.722005e-4 months <br /> on June 18 and has since operated continuously at a gross output of 63 We (gross). The off-gas activity was averaging approximately 1,000 pCi/s at the end of this report period.
II.
EFFLUENT AND ENVIRONMENTAL MONITORING l
A.
Introduction and Conclusions 1.
Introduction Releases of radioactive material both to the atmosphere and Lake Michigan from January 1 to June 30, 1972 were well within the facility-licensed limits and the Commission's regulations, particularly Title 10, Code 'of Federal Regulations, Part 20.
Environmental levels of radioactiv-ity as found in the vicinity of the plant were composed almost entirely of
-naturally occurring radioactive materials. Only in the vicinity.of the circulating water discharge canal was radioactive material of plant origin fcund. These materials occurred primarily in aquatic organisms. These levels of radioactive materials, however, were extremely low and pose no 4
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threat to the health and safety of the public. Further, these levels of radioactive material found in the resident biological emity are con-sistent with levels found in previous years and show no upward trend.
2.
Atmospheric Effluent Gaseous releases to the atmosphere totaled 169,500 curies of fission and activation gases. This corresponds to 1.07% of the licensed technical specification limit of 1 Ci/s. Particulate releases totaled 0.0072 curie or 0.04% of the licensed limit while halogen releases totaled 0.08h curie or 0.44% of the licensed limit. Tritium releases totaled 29.8
-E curies or 5 1 x 10 % of a limit based upon meteorological dispersion to the point of maxi =m ground concentration.
The calculated radiation dose at the site boundary resulting from these releases was 5.h millirems. The integrated dose to the popu-lation out to 50 miles was 4.9 man-Rems.
.3 Liquid waste Releases Liquid waste releases totaled 0 558 curie of radioactive mate-rial. This release corresponds to 0.66% of technical specifications limits. Additionally 9 3 curies of tritium were released corresponding to 0.006% of 10 CFR 20 permissible concentrations in the discharge canal.
Population doses based upon drinking water from the Charlevoix municipal system was 0.0012 man-Rem and total Lake Michigan drinking water consumption population dose'was 0.04 man-Rem..The consu=ption of all of the Lake' Michigan fish harvested resulted in a population dose of 0.073 man-Res.
4.
Solid Wastes A total of 1,039,542 curies of radioactive material was shipped off site during the period covered by this report. Out of this total, 1,035,000 curies were irradiated cobalt and 4,542 curies were solid waste.
See Appendix C.
B.' Atmospheric Effluent and Environmental Su=ary
-1.
Effluent Calculational Methods A ' sample of off-gas is 'obtained weekly during power operation and analyzed by.ga=a spectrometry for *six noble gas radionuclides.
Based upon the mixture of the six nuclides, a stack release rate,.which
'The six nuclides are: h-85m, -87, -88. and Xe-133, -135 'and -138.
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6 includes a total of 22 noble gas radionuclides is determined. This 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 distribution 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 Arpendix A.
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.
Determination of release rates in this manner assumes radioactivity is continually being deposited throughout the week on the filters and, hence, a decay correction to the time of analysis is applied, depending on the half-life of the nuclide observed.
Gross alpha measurements on the particulate filter revealed that the release of alpha-emitting nuclides is about 2 to 14 orders of magnitude lower than beta-emitting nuclides.
Tritium releases to the atmosphere are calculated, based upon measurements made in the primary coolant and containment air and using identical concentrations for all releases as follows:
a.
Off-Gas - A flow rate of 10 cfm containing 100% relative l
humidity and 90% radiolytic gas by volume both at primary coolant tritium to hydrogen ratio to determine-tritium releases both in vapor and molecular form. The increase in tritium released to the atmosphere over last year is
. primarily due to the increased tritium concentration in the primary coolant.
This in turn-is due.to the increased recycling of liquid radioactive waste back'into the primary coolant system.
- b. - Turbine Sealing Steam - The measured flow rate at 100%
c.
Containment Ventilation - The measured flow rate and mea-sured containment building tritium concentration.
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7 The results of these calculations are also shown in Appendix A.
2.
Environmental Methods and Data In order to predict potential radiation deses resulting from the above releases, environmental transport and uptake factors must be known.
A confirmation of these calculated doses is attempted then by measuring levels of radioactive materials in the plant's environmental surveillance program. In previous reports for the Big Rock Point Plant, the average yearly meteorological relationship between release rate and downwind concentration of radionuclides at distt.nces from the plant up to 50 miles has been used. Using a typical equilibrium mixture of noble gas release 3
after a 30-minute decay, a concentraticn of 1.4 x 10 pCi/cm delivers in air an exposure rate of 1 millirem per hour using semi-infinite cloud
-1 3
geometry. FromthisandtherelationshipX/4=3.45x10 s/cm, the release rate required to produce a dose of 500 millirems per year at the point of maximum ground concentration was 2.4 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 10 ) x (MFC) which produces a con.
centration at the point of maximum ground concentration some 2400 times below maximum permissible concentration allowed in Title 10 CFR 20 for I-131 using the previous meteorological relationship.
3 TheX/Qvalueof3.45x10 s/cm 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% of the time per Section 915.11 of the FHSR.
UsingthemeteorologicaldataintheFHSRX/Qvalues,foreachsector, have now been calculated which account for the amount of time the differ-ent stability classes exist and the wind is blowing in each sector. At the site boundary, a plume from the plant has not yet reached ground level.
Therefore, any dose received from plant releases at the site boundary would be a shine dose from an overhead finite cloud. Per "Meterology and Atomic Energy - 1968," using a diffusion mixture, the release rate required to deliver 500 millirems per year at the site boundary is 0.66 Ci/s in the critical sector which is Sector 4 (see Appendix D). For an equilibrium mixture, on the other hand, the release rate required to deliver 500 i
millirems per year at the site boundary is 1.05 Ci/s.
0 i
The environmental surveillance program includes continuous sampling of air for particulate and halogen activity at seven locations including background sample locations at Traverse City and Boyne City, Michigan, about 50 miles south-southwest and 20 miles southeast of the plant, respectively, to detemine increased concentrations, if any, 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 environcent. Any increase in direct dose at the closer stationc can then be attributed to be of plant origin due to direct radiation from the plume.
3 Environmental Dose Calculations A computer model is now used to calculate radiation dose re-sulting from plant releases of noble gases. The integrated population dose, out to 50 miles, for the first six months of 1972 is shown on the following page. The computer model utilizes the following:
X/Q values for the five sectors are averaged over both a.
stability class and wind frequency.
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 summation of the individual nuclide contributions.
c.
The 1972 population is estimated from the 1970 Census of 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 estimated 1972 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 release 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 plume before it reaches the different annuli.
9
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Calculated Fadiation Doses from Gaseous Relesse.
January 1,19"(2 to June 30, 1972 SIETOR 1
2 3
L 5
Total 1-2 Population 13 Th 0
10 0
97 Population Dose 1 560x10-2 h.167x10-2 0.0 1.036x10-2 0.0 6.763x10-2 2-3 261 266 0
50 72 6L9 Population.
2.033x10-1 1.089x10-1 0.0 3 757x10-2 5,tupxio-2 g,og1xto-1 Population Dose 34 Population 555 392 0
47 57 1051 Population Dose 3 092x10-1 1.26hx10-1 0.0 2 753x10-2 3 2f+8x10-2 L.956x10-1 k-5 Population 3300 712 0
102 0
L114 Population Dose 1 366 1 767x10-1 0.0 4.725x10-2 0.0 1 5d9 5-10 Population 2074 24 0
527 0
2625 Population Dose
'3 806x10-1 2 747x10-3 0.0 1.lo2x10-1 0.0 5 025x10-1 10-20 Population 8868 390 737 13,938 323 24,2t6 Population Dose 3 983x10-1 1.080x10-2 h.h23x10-2 8.231x10-1 1.605x10-2 1.292 20-30 Population 9523 3kS8 1895 4564 323 19,763 Population Dose 1.18x10-1 1 71x10-2 2.1x10-2 7 8x10-2 L.3x10-3 2 33hx10-1 3'> LO Population 22, h72 LO27 2P,77 L783 0
34,161 Population Dose 1.09x10-1 1.15x10-2 1.87x10-2 3 21x10-2 o,o 1,71xic-<
LO-50 Population 40,251 8770 5795 11,941 0
66,757 Population Dose 9 53x10-2 1.2hx10-2 1.86x10-2 3 9x10-2 0.0 1.66x10-1 0-50 Population 87,319 18,113 11,304 35,7h3 775 153,25L Population Doce 2.9) 5 111x10-1 1.082x10-1 1.219 1.12hx10-1 L.9L9 Cite poundary
. Dose ( rem /6 mos) 4.614x10-3 2 9%x10-3 5 7LLx10-3 5 434x10-3 5 229x10-3
10 Doses from particulate iodine and tritium releases as shown j
in Appendix A were negligible ccmpared to that received from noble gases due to the conservative limits in the plant technical specifications and the absence of any significant milk food chain in the area affected by the plant.
.In order to obtain greater sensitivity of measurement a ecm-parative program of film vs thermoluminescent dosimeters (TLD) was started in late 1971. The program consists of placing a film and TLD side by ride at each monitoring station for a one-month exposure period. Certain aspects t
of the TLD program, mainly the handling and shipping procedures, have re-sulted in erratic readings. These difficulties are expected to be resolved soon, at which time the results of these dosimeter analyses will be in-cluded in the environmental surveillance program.
Air samples gathered continuously and analyzed weekly at the stations shown in Appendix D showed no difference 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 poten-tial concentration of radioactive materials resulting from plant operations.
From the known meteorological dispersion conditicas, the following maxi =um concentratiens ean be calculated:
's/cm) 3
~1 Particulates (January)
(1.2 Wi/s)' x (0.0012) x (5.0 x 10
=. 7.2 x lo" 7/pCi/cm3
-1 3
Halogens-(February)
(1.2 pCi/s) x (0.0094 x (5.0 x 10 s/cm)
-16 3
5.64 x 10 pCi/cm
=
These-compare to the minimum detectable activity values and normal background concentrations as follows:
Maximum Calculated Minimum Detectable Normal Background 3
Release Concentration uCi/cm Activity uCi/cm3 Activity uCi/cm3 1"
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Particulate 72xlo l x 10 7x10
. Halogen 5.6 x lo~D 2 x 10-13 Hence,. the negative data obtained in the program was expected.
-C.
Liquid ECT1uent and Environmental Summary 1.
Effluent Calculational Methods The release pathway to' Lake Michigan for all liquid effluents is through the plant's condenser circulating water: discharge canal. A Y
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flow rate of k8,000-52,000 gpm dilution for liquid effluents is obtained through the use of the condenser circulating water pumps, two at 2h,000 gpm each, and house service water pumps, two at i, om each.
Each collected tank of liquid is sa=ple d, analyzed for radio-active content, and discharged at a controlled rate to assure that per-missible concentrations are not exceeded in the canal prior to dilutien in Lake Michigan during the time of discharge. Each sa ple is analyzed by ga==a spectrometry to identify as many of the co=ponent nuclides as possible.
(See Appendix B for results.) Permissible concentrations in the canal are determined from the following:
Ci I
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MFCi 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=ma spectrometry but measured by a gross beta analysis are presumed to be Sr-90 and released on that basis. Sa=ples of the batches are then sent to the radiological environ-mental contractor and analyzed for Sr-90 and Sr-89 From concentrations of Sr-90 and Sr-89 found in the batches, the total curies released of these two isotopr. is calculated and used in calculating the percent of applicable limit in pendix B.
The re=aining unidentified isotopes are assigned an
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.0 pCi/ml per 10 CFR 20.
Tritium measurements are now being made on most batches released.
2.
Environmental Methods and Data In _ order to predict potential ~ radiation doses _ resulting from the above releases, environmental transport and uptake factors must be known.. A confirmation of these calculated doses is then attempted by
. measuring levels of radioactive materials in the plant's environ = ental -
radiation surveillance pro 6 ram. At the_ Big Rock Point Plant, daily co=-
. posite 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 radionctive content. These results are shown in~ Appendix D.
Additional aquatic samples are taken 1
and an'ilyzed during the summer growing season and these results are also tabulated in Appendix D.
l 12 4
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Based upon the release of 0.558 curie of radioactive material (less tritium) which results in an annual average concentration in the discharge canal of 1.1 x 10- uCi/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 er about 2 times lover t. ban the as-erage concentration discharged. The results shown plotted in Appendix D
-9 indicate an average of about 9 x 10 uCi/ml for the year.
The nearest municipal drinking water supply intake is located in Charlevoix,' Michigan which is generally upstrea= of the prevailing 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 ficv all year in calculated in the next section of this report. 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 Report No 9," by John C. Ayers, 1961.
In addition, the population dose is calculated to the entire population which receives its drinking water from Lake Michigan, based on a uniform concentration, resulting fro = plant releases, throughout Lake Michigan.'- Also, radiation dose to human populations can occur as a result of plant releases through the consu=ption <>f fish caught in Lake Michigan.
3.
Environmental Dose Calcalations
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-Utilizing the measured values.of radionuclides released as shown in Appendix B, the following formula, and the standard man model, drinking water doses can be calculated as'follows:
f Ci I g
1 D, =
I ng se RemM
{ MPC ).
1 D,is.the. individual dose in Rem /yr, where:
l Ci is the averaSe -concentration in Lake Michigan of the individual nuclides measured, in pCi/ml, MPC.is the' concentration of each nuclide measured required to
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produce the limiting' dose at continuous intake in pCi/ml and Limiting' Dose.is the dose produced at continuous exposure'to
._MPC concentrations.
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- u In calculating ingestion dose from the consumption 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 water 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 consuming fish, are per the previous formula wh'.le integrated popula-
~ tion doses.in man-Rem are calculated utilizing the following parameters:
For drinking water, the individual doses are summed over a.
'the entire population that receives its drinking water from Lake Michigan
~ vith discharge canal flow appropriately mixed with the lake. This 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 3500 people.
- c. "For fish consumption, the average concentr.ation in Lake I
Michigan wat'er, resulting from plant releases, is used with a bioaccumula-tion factor-to determine the average concentration in fish, i-d..
Fish do act reside continuously in the discharge canal but f
migrate. This can be. seen in the following table -which compares the fish consumption dose based on the discharge icanal water concentration and the
. appropriate reconcentration factors -to the fish consumption dose calculated from actu'al concentrations in fish caught in or near the discharge canal.
l l
- 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 ofl Radio-
) isotopes-in the Food Web," May;1972..
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- A'
14
" Fish Consum ^. ion Fish Consumption Average Con-Average Con-Dose Calculated Dose Based on centration in centration in From Discharge Canal Ceccentration in Discharge Canal Sampled Fish Concentration Sampled Fish Isotope (uCi/ml)
(uCi/g) mrem /Yr mRe=/Yr
-9 19 I-131 1.0 x 10
-9 Cs-134 2 5 x 10 7.h
-9
~I Cs-137 2.6 x 10 2.h x 10 3.5 0.1h
-10 Co-58 2.1 x 10 2.4
-9 Co-60 1.2 x 10 0.h6
-10
-7 Zn-65 4.6 x'10 3.h x 10 0.046 0.0039
-11 0.054 Sr-89 9 1 x 10
-1 3.0 Sr-90 6.1 x 10 Others 2.1 x 10 0.062
~
The fish consumption dose calculated from discharge canal concen-
.trations is 10 to 20 times as large as the fish consumption dose calculated from actual concentrations in fish. -
As a measure of total environmental' impactL the radioactive liquid releases frou the plant ere averaged over the entire lake and then used.to determine the population dose from fish cr.ught throughout the entire lake and total water consumed from the lake.
Both of the dose calculations are conservative in that:
. Equilibrium is not' obtained in the human body for most a.
isotopes released.
b.
No _ credit is taken for precipitation and deposit in sediment or uptake by life forms other than fish-which are seen to occur by 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 1972.
c
.m m
.m..
5
CAlft' LATED RADIATION IrSES From Liquid Effluents - Population Drinking Vater Dcse January 1,1772 to June 30,1?72 Avg Concentrati Population Dosa Critical Curies in I,6ke Michigsn (i!
Pooulation Dose (
Charlevoix,Mich(3) 1 i
C/E Vector Isot Se MPC Orgsn Releases (uCi/ml) i i
tr. Rem /Yr Man-Rem Min-Fem
~7
~N 4
~3 Water I-131 3 x 10 Thyroid 0.052 1.1 x 10 3.7 x 10~
1.0 x 10 0.10 7.5 x 10
~1
~9
~7 Cs-134
'9 x 10" Whole Body 0.126 2.6 x 10 2.6 x 10 7.1 x 10 0.0071 3 9 x 10~
-5 M
Cs-137 2 x 10 Whole Body 0.133 2.8 x 10 1.4 x 10-9 3 8 x 10~7 0.0038 2.8 x 10-0 Co-58 1 x 10~
GI Tract 0.0106 2.1 x 10~13 1.1 x 10~9 3.2 x 10 0.032 1.4 x 10-3
~5
~1 Co 60 3 x 10 GI Tract 0.0603 1 3 x 10 4.3 x 10~1 3.5 x 10-7 0.0035 2.6 x 10" Zn-65 1 x 10 Whole Body 0.0232 4.8 x 10 4.8 x 10 1.3 x 10-8 0.00013 1.0 x 10~5
-15
~11
-16
~10
-7 Sr-89 3 x 10~
Bene 0.00L6 9 1 x 10 1.6 x 10 3 x 10 0.003 1.3 x 10
-16
~9
-6
-3
~
Sr-90 3 x 10 Bone 0.0031 6.5 x 10 2.2 x 10 9.8 x 10 o,o93 7,g, 10
-3
~1
- I H-3 3 x 10 Whole Body 9 26 1.9 x 10 6 3 x 10~
1.7 x 10 0.0017 1.3 x 10~'
-6 Others 3 x 10 Whole Body 0.027 3.1 x lo-1.0 x 10~
2.7 x 10 0.027 3.9 x 10 Total Whole Bnly 0.04 1.2 x 10~3 Thyroid 0.10 7.5 x 10~3 GI Tract 0.036 1.7 x 10-3
~3 Bone 0.10 7.5 x 10 (1) Average concentration in Lake Michig =curiesreleased/volumeofLakeMichigan.
Volume of Lake Michigrtn % 4.8 x 10 liters.
Population taking its drinking water from Lake Michigan is approximately 10,000,000 people with 7,000,000 in the Chicago area.
(3)Using average concentration in discharge canal diluted by 800.
10 CFR 20 MPC for unknown mixture with certain isotopes not present.
y
}This compares to a background and. medical radiation dose of 0.215 Rem /yr/ person or 2.15 x 10 man-rem for the population taking its drinking water from Lake Michigu.
-CAIfULATED RADIATION DoGEIi a
Fren Liquid Effluent - Fish Constunption Dose
. January 1 '1972 to June 30, 1972 Avg Concentration Avg Concentration Population Doze (3) g (1)
Critical Biomecumua} ion in Lake Michigan' in Fish fi i) MUi Vector Isotope i
organ Factor 2 pC1/ml uCi/c mrem /Yr Man-Rem
/
~5
-1
-12 2.2 x lo'N o.125 Fish I-131 1.3 x 10 Thyroid'
'500 1.1 x 10 5 5 x 10
~5
.Cs-134
'4.0 x lo-Whole Body 2360 2.6 x lo~l 6.1 x 10-11 7.8 x 10 0.045
-5 Cs-137 8.8 x'1(
Whole Body 2360 2.8 x 10-1 '
6.6 x 10-11 3.8 x 10 0.022
-15
-13
-5 0.014
' Co-58_
k.4 x lo-5 at Tract
'330 2.1 x 10 7.x 10 2.4 x 10
~1
-12
-6 Co-60
'1 3 x'10 CI Tract 330 1.3 x 10 4.3 x 10 5,1g g,gg3 L Whole Body 900 6 8 x 10~15' 4.2 x 10-12 g,g,yg-7 g,ggg3 Zn-65' 44xid'3
-16 7,19-14 3,g,gg g,ggg3
-7 sr'-89 1.3'x 10
- Bone 80 91x10 Sr-90 1 3 x lo'I ' Bone 80 6.5 x 10 5 x 10~1k 3.2 x 10 0.019
-1
-5
-N
-12
-6 others.
'1 3 x 10
'Whole Body 80 3 1 x lo 2.4 x 10 9 2 x 10 g,gg3g Total Whole Body 1 3 x lo 0.073 Thyroid 2.2 x 10~
0.125 GI Tract 5 5 x 10 0.0033
-5 Bone 3 3 x 10 0.025 I>I) Maximum concentration for fish = MFC,* (50/2200).
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."
Using 23,873,689 pounds of fish harvested from Lake Michigan in 1970. This number includes both commercial and sports catches as shown
~
in Appent'ix D minus alewives which are not generally c.onsumed.
5
}This compares to an average background and medical radiation dose of 0.215 Rem /yr/ person or 13 x 10 man-rem for the population necessary w*
to consume the Lake Michigan fish catch at a rate of 50 g/ day / person.
c
17 III. PRINCIPAL MAINTENANCE PERFORMED A.
Reactor During the refueling outage in March, four in-core detector assemblies in Positions 13,14,15 and 16 were removed from the reactor vessel and replaced with four new in-core detector assemblies.
The No 6 head bolt was removed from the reactor vessel flange for ultrasonic testing and was replaced with a spare head bolt. No deficiencies were noted during the test.
B.
Primary System Weld Inspection An ultrasonic examination of primary system piping velds was performed by Southwest Research Institute personnel. Preliminary results were satisfactory with no defective velds noted.
The inspected welds were as follows:
1.
Four circumferential pipe velds in the Main Steam System.
2.
Four circumferential pipe velds in the Feed-Water System.
3.
Four circumferential pipe velds in the Emergency Condenser System.
h.
Thirteen circumferential pipe veld? in the Clean-Up System.
5 Three circumferential pipe -velds in the Shutdown System.
6.
Eight circumferential pipe velds in the Main Recirculation System.
1 7
One circumferential pipe veld in the Core Spray System.
8.
Three circumferential pip' velds in the Redundant Core Spray System.
9 Six circumferential pipe velds in the Poisen System.
l 10.
Eighteen circumferential pipe velds at the newly installed l
Clean-Up System heat exchangers.
13. Eighteen nozzles of the newly installed Clean-Up System heat exchargers.
12.
Two velds of transition at Clean-Up System heat exchangers.
- 13. Ten Clean-Up System heat exchanger velded supports.
- 14. Two "J" welds attackin6 tontrol rod drive housings to stub tubes for Drives B-5 and C-5 The insulation removal and the ' installation of snap-on insulation j,
- was performed by Consumers -Power Company maintenance personnel. With the.
L
4 18 i
f use of snap-on insulation, reinspection of this piping should result in considerable lower personnel exposure in the future.
C.
Control Rod Drive System Prior to the refueling outage, control rod Drives C b and D-6 developed excessive cooling vater leakage necessitating change of the drive 0-ring scals. During the refueling outage, control rod Drives B-5, C-3 and C-5 were removed and replaced with reconditioned drives. The "J" welds on the B-5 and C-5 drive housings were examined ultrasonically and found satisfactory. New style silver-plated inconel 0-rings (manufactured by United Aircrafts Products) were installed for flange sealing in Drives B-1, B-3, B-5, C-3, C-5 and F-2.
However, after the refueling outage and on plant start-up, Drives B-5, C-3 and F-2 developed excessive leakage at the flanges. Because of high temperature, B-5 was removed and replaced with a rebuilt drive. The silver-plated inconel 0-rings apparently did not compress enough when the flange bolts were torqued and had to be re-f placed with the old style Teflon type. The remaining three drives with the inconel 0-rings (B-1, B-3 and C-5) were retorqued and have perf ;med satis-factorily to date. The silver-plated inconel 0-rings were installed because of better flev characteristics at high temperatures. Clearance and align-ment measurements were taken of the control rod drive support str _cture modules relative to the rod drives. The support modules are within clear-i ance specifications.
D.
Reactor Recirculating Water Pumps Both No 1 and No 2 recirculating pump seals were replaced during the outage because of unequal pressure distribution between the two seals on the No 1 pump and erratic pressure distribution across the seals on the No 2 pump.
After approximately three days of service and another shutdown to replace the No-1 pump 3/h-inch heat exchanger, the No 1 pump seal had to' be. replaced because of a lack of seal controlled leak off flow rate.
Inspection revealed thst the seal was plugged, preventing cooling water flow from reaching the upper seal. The.3/k-inch heat exchanger had to be i
replaced because of a leaking tube' coil.
t' 1-
---+w
,m-_.g>,
19 E.
Penetration Welds The annual inspection (visual) of all containment sphere penetra-tion velds as required by the Technical Specifications was conducted with no discrepancies noted.
F.
Clean-Up System and Heat Exchanger Replacecent Leaking drain Valves CU-118 and CU-120 on the Clean-Up System vere disassembled and inspected.
Valve discs and seats were relapped and the valves repacked, correcting water leakage from the primary system.
New replacement Clean-Up System heat exchangers were installed by Livscy Corporation. It was during the installation of the heat ex-changers that manufacturing defects were found in several nozzle-to-shell velds and in the interconnecting piping velds. These were repaired on site to the applicable codes by the manufacturer (Southwestern Engineerin6 Company) and an outside contractor (Livsey Corporation) under *he direc-tion of our Construction Department.
G.
Turbine Generator Because of a faulty trip mechanism which initiated a turbine trip, the turbine trip solenoid was leplaced with a spare unit incor-porating a heavy-duty, continuous use type trip solenoid.
Linkages on the turbine initial-pressure regulator (IPR) were IFR control adjusted in an effort to control the turbine at lover loads.
of the turbine can now be maintained down to approximately 10 MWe. Previous IPR control could not be maintained below 30 MWe.
Six sections of pipe and seven pipe fittings in the turbine extraction drains were replaced as a result of a thin vall (erosion) condition. Wall thicknesses were determined by radiography. This is the third year of a program for replacing Schedule h0 piping with Schedule 80 piping in the turbine stage drains. The program has greatly improved turbine reliability.
Twenty-two main condenser tubes were plugged and additional stainless steel wire mesh was installed over the condenser tubes in the area polished by steam impingement. The mesh previously installed appears to have retarded surface erosion of the tubes effectively.
(
r 20 r
H.
Emergency Condenser System A new d-c motor was installed on motor-operated Valve MD-7063 which is on the outlet of the No F loop of the e=ergency condenser. The removed motor has been sent to iba factory for complete overhaul, including revinding, and vill be available as a ready spare if needed. This motor had previously been revound twice at a local shop because of accidental shorting across the vindings.
I.
Energency Diesel Generator A new emergency diesel generator water pump was installed, re-placing one that was worn and lacked sufficient priming capacity. However, the pumping capacity was net sufficient and the old pu=p was rebuilt and reinstalled. A new, larger, replacement pump will be installed during a
- convenient outage following its arrival.
IV.
CHANGES, TESTS AND EXPERIMENTS PERF0FJED PURSUANT TO 10 CFR 50.59(s)
A.
Facilit;v Chances No C-18k - Plant Security Lightine and Fencine - Four-hundred-watt' pole lighi: with individual sensing units were added at the well-house, screenhouse, radwaste area and along the vest perimeter fence to
~
increase plant security. Seven-foot chain link iencing with three strands of barbed wire was installed,-enclosing the well house and water storage tank.. Fencing was also installed at the public information center, in-creasing a 30-foot section of fence to a height of 16 feet.
No C-166 - Instrument Niring Change - The power supply for the dissolved oxygen analyzer was changed from the acid and caustic power 1
)
supply breaker (lighting' Circuit 3L-26) to the condensate demineralizer control panel feed (instrument and centrol Circuit 1Y-29) to provide Lgreater-source' reliability.
-No C-lT1 - Instrument Wiring Change '- The power supply for the containment sphere continuous air monitor was changed from lighting Circuit 12L-27 to a spare instrumentation circuit (instrument and control Circuit -
2k-k),to provide. greater source reliability.
No C-172 - Public Address System - The reactor refueling deck area public address systen speaker and associated amplifier were relocated.
- from.the north side of the. area to the east side. This eliminated the
- i
21 reflection (feedback) from the steam drum vall and provided a more even coverage over the fuel pool refueling area.
No C-16h - Containment Sphere Pipevay Dev Cell - The change made to this system consisted of the addition of a reference dev cell, to enable an increase in dev point temperature from leaks originating in the contain-ment sphere pipevay to be referenced against the ambient dev point temper-ature in the containment sphere.
No C-17h - Containment Sphere Test Instrumentation - Installa-tion of a new ambient dew point temperature recorder was completed during the refueling outage. This addition vill provide ambient and dev point temperatures at six locations throughout the plant. Although the primary purpose of this recorder is for biannual containment sphere testing, two locations provide backup electrical readout of dev point temperature for the two pneumatic systems presently in use. All points are connected to a 12-point recorder located in the control room.
No C-179 - Turbine Bleeder Trip Valves - The pneumatic control piping on the bleeder trip valve solenoid valves was changed to improve valve operation-(BTV-kh50 and hkS1). This involves reversal of the feed-vater heate.- high level and-the test solenoid valves and the addition
~
of a check-valve in each bleeder trip valve scheme (high pressure and intermediate pressure)-to provide rapid air dump following turbine trip.
Also, the solenoid valves were rebuilt and relocated from a position neer the bleeder. trip valves to.the vall of'the pipe tunnel in an attempt to prevent temperature - deterioration of the - solenoid valve diaphragts.
No C-178
- Al'r Ejector Off-Gas Drain Line - This change involved installation of a solenoid valve in the drain line to provide closure of
'the drain line wher,the off-gas isolation valve c'oses. Prior to this
~
addition, the. only isolation provided on this drain line was la 26-inch loop-seal.
No C-161 - Stack-Gas Sample System - This change involved in-stallation of a three-way solenoid in the scintillation detector inlet sample -line to close-the sample 111ne during purge; thus providing com- -
v.
- plete purge control from the c
>1-room without requiring local manipu-
- lation of the valving.
l c
22 No C-180 - Liquid Poison System - New low temperature primers a
i were installed in all explosive valves.
(Reference Tech Spec Change Request No 33 dated July 20,1972.)
No C-176 - Emergency Diesel Generator Coolant Pump - This change involved fabrication of a flexible spool piece so the emergency diesel i
generator coolant pu=p primary water could be obtained from either the plant domestic water system or the plant fire water system in case house service water was not available.
No C-165 - Redundant Core Spray System Pressure Relief Valve -
This change involved installation of a pressure relief valve in the core spray system to protect from overpressurizing the core spray piping. Valve size is 7.6 gpm at 140 psig.
No C-177 - Reactor Clean-Up Systean - This change involved installation of a manual 2-inch bypass valve on the 6-inch outlet valve of the component cooling water on the nonregenerative heat exchanger.
This valve allows better Vater flow control and, thereby, better temper-ature control, increasing efficiency. The additional valve vill also increase lifetime of the 6-inch outlet valve.
B.
Tests A temperature coefficient test was conducted in f/ay prior to
' power operation with the newly-loaded core. Test data indicated that the temperature coefficient turned negative at Ih2 F after adding 104 worth of reactivity.
b The containment sphere leak rate test was conducted successfully.
j.
A complete report viil.be submitted at a later date.
V.
PERIODIC TESTING PERFORMED AS REQUIRED BY THE TECHNICAL SPECIFICATIONS-i 7
The following tabulation shows the required frequency of. testing l
- plus the testing date of the systems or functions which may be periodically tested per Technical-Specifications:
i-lt t
i j
l l
v 23 System or Function
- Frequency of Undergoing Test Routine Tests Dates Tested Control Rod Drives Continuous withdrawal and inser-Each major refueling and 5/h/72 tion of each drive over its at least once every six stroke with normal hydraulic conths during periods of system pressure. Minimu= with-power operation.
drawal time shall be 23 seconds.
Withdrawal of each drive, stop-Each major refuelin6 and 5/h/72
-ping at each locking position to at least once every six check latching and unlatching months during periods of operations and the functioning power operation, of the position indication system.
Scra= of each drive from the Each major refueling and 5/h/72 fully withdrawn position. Maxi-at least once every six mu= scram time from system trip months during periods of to 90% of insertion shall not power operation.
exceed 2 5 seconds.
Insertion of each drive over Each major refueling but 5/3/72 its entire stroke with reduced not less than once a hydraulic system pressure to year.
determine that drive friction is normal.
Control Rod' Interlocks Rod withdrawal blocked when any Each major refueling but 5/1/72 two accumulators are at a pres-not less frequently than sure below 700 psig.
once every twelve months.
Rod withdrawal blocked when two Each major refueling but 5/2/72
- of three power range channels not'less frequently than read belov 5% on 0-125% scales.
once every twelve months.
.(or below 2% on.their 0 h0%
scales) when reactor. power is above the minimum operating range of these channels.
Rod withdrawal blocked when Each major refueling but 5/1/72 scram dump tank is bypassed.
not less than once every twelve months.
Rod withdrawal blocked when-Each major refueling but 5/1/72 mode selector switch is in not less "requently than-shutdown position..
-once every twelve months.
' Reference 2/23/72 letter to the AEC.
(
1
2L Syste= or Punction
- Frequency of Undergoing Test Routine Tests Dates Tested Other Liquid poisen syste= co:-
Two conths or less.
2/1k/72, 3/6/72 and 5/2/72 ponent check.
Post-incident spray syste=
At each major refueling 5/11/72 autc=atic control operation.
shutdown but not less frequently than once a year.
Core spray system trip circuit.
Not less frequently than 5/11/72 cnce every twelve tenths.
EmerEency condenser trip Not less frequently than 5/11/72 circuits.
once every twelve months.
Containment Contain=ent sphere access air Once every six conths 3/2/72 locks and vent valves, leakage or less.
rate.
Isolation valve operability At least once every h/k/72 and leak tests.
twelve conths.
Isolation valve controls and Approxinately quarterly.
1/25/72 and L/4/72 instrumentation tests.
Penetration inspection.
At least once every 3/30/72 twelve conths.
Integrated leak test.
Once every two years.
h/5/72 The following instrument checks and calibrations were perfor=e'i at least once a centh:
1.
Reactor safety system checks not requiring plant shutdo.n.
2.
Air ejector off-gas monitor.
3.
Stack-gas monitor calibration.
h.
E=ergency condenser vent monitor.
5 Process monitor.
6.
Area monitoring system.
' Reference 2/23/T2 letter to the AEC.
4
25 VI.
RADIOACTIVE LEVELS IN PRUCIPAL FLUID SYSTEMS 4
Mininum Averare Maximum A.
Primary Coolant Reactor Water Filtrate ("}
-2
-1
-1 pCi/ml k.37 x 10 2.18 x 10 5 1 x 10 Reactor Water Crud (c)
-1
~
pCi/ml/ Turbidity Unit 2.04 x 10 2.63 x 10 1.16 Iodine Activity (b)
~
1.1 x 10 15
~
pCi/ml 1 x 10 J
B.
Reactor Cooling Water System Reactor Cooling Water ")
l
-3
-2
-2 pCi/ml h.35 x 10 1.h5 x 10 3,35 x yg t
C.
Spent Fuel Pool Fuel Storage Pool (*}
-1
~
~
pCi/ml.
8.7 x 10 1.17 x 10 7 3 x 10 Fuel Pool Ionine(b)
-7
-3 pCi/ml 8 x 10 3 x 10" 6 x 10 (a)A counter efficiency based on a decay schese consisting of one gn=ma photen per disintegration at 0.662 MeV used to convert count rate to microcuries.
All count rates were taken at two hours after sampling.
(b) Based on efficiency of Iodine 131 two hours after sa:pling.
" Based on APHA turbidity units and 500 ml of filtered sample.
CONSILERS PDFER COMPANY n/h w<L By
/
Hucleat Licensing Administrdtor Date: August 30, 1972 Sworn and subscribed.to before me this 30th-day of August 1972.
lI jQ' rt. '
- )
.<D
'c cr /
<f te.) s) 'V Notary Public, Jackson County, Michigan W commission expires June 20, 1976
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4 APPINDIX B Consumers Power Company i
Big Rock Point Plant Docket: 50-155 -
Radioactive Liquid Releases tinits J.amary February March April May J'me_
Total Total Radioactivity Released (Execpt Alphah)DissolvedCasesand Triti C1-0.042
. 0.179.
0.087 0.25 0.558 5
Volume of Waste Discharged Liters 2.28 x 10 1.48 x 10 8.42 x 10 8.83 x 10 3.43x10f A' verage Concentrat$ of Waste Prior to Releasel -
pCi/mi 1.84 x 10'3 1.21 x 10-3 1,o3,19 2.3 x 10~3 1.63 x lo*3
-3 Volme of Circulating Discharge 9
9 9
8.18 x 10 8.78 x 10 8.5 x 1r 5.06 x 10 9
9 10 kater Liters
'8.79 x 10 7.63 x 10 8.73 x 10 Average Concentration Beleased (Exced "
-8
) Dissolved Cases and
.pC1/ml 5.5 x 10~9 2.05 x 10~0 1.06 x 10 2.85 x 10'O Triti 1.1 x 10 Alpha
. Maxisma Concentration cept Tritium 6.3 x 10"I 5.9 x 10-7 4.9 x 10-7 9 9 x 10~7 9 9 x 10-7 and Dissolved cases) pC1/mi I) 0.84 0 74 0.4 2.0 OM Percent of Applicable Limit
=
Mixture MPC 301/m1 6.6 x 10'I 2.8 x 10 2.6 x 10 1.4 x 10~
1.7 x 10 0.41 6.19 0.44 2.22 9.26 Triti*un Released -
C1 4
- 7 Average Tritium Concentration 4
7.1 x 10-7 5.4 x 10 2.5 x 10*7 1.8 x 10 Released pC1/ml 5.4 x 10 3
Isotopes 0.0031 0.011 0.0055 0.0 32 0.0516
.1-131 C1 0.024 0.074 0.019 0.009 0.126
- Cs-134 C1 0.034 0.023 0.076 0.133 Cs-137 C1 0.0042 0.042 0.006 0.0081 0.0603 Co-60 C1 0.016 0.0072
,g o.0232 4.6 x 10'3 2n 65 C1 3.4 x 10'3 2.1 x 10'"g 9.7 x 10,g
-3 sr-89
. Ci '
2 3 x 10 k.4 x 10 3.4 x 10 0.01M
~3 3 1 x 10 Sr-90 Ci 0.0106 Co-58 Ci 0.412 Total taentified Released Radioactivity 74 Percent of Total Identified
( Dissolved gases were below detectable limits, alpha emitters were 2 to 6 orders of magnitude less than beta emitters.
i Average concentration of wasta prior to release a pC1 released for the month /ml of waste discharged.
0 (3) Average concentration released a pCi released for the auxtth/ml of circulating discharge.
Maximum cone *ntration during actual. batch release.
d.
Percent of applicable limit = ***#*"
- mixture MIC i
Appendix C Transfer of Radioactive Material Shipment Transfer No.
Date From Transfer To Radioactive Material 249
-1/14/72 DPR-6 G.E., San Jose, Calif 3 In-Core Chambers O.3 mci SNM-54 250 1/28/72' DPR-6 NRL, Washington, D.C.
Irradiated Vessel Specimens 109.6 Ci 8-1393-2 A-66 251 1/26/72 DPR-6 Battelle Columbus, Ohio 3 Irradiated Fuel Rods 3324 Ci
-SNM-7
'252 2/1/72 DPR-6
.Battelle Columbus, Ohio 1 Irradiated Fuel Rod 1108 Ci SNM 253' 2/4/72
~DPR-6 Isotopes Inc., I;ew Jersey Reactor Water and 0.E mci 29-55-6
. Condensate Samples 254 2/7/72 DPR-6 G.E. Vul. 0017-60 Feed-Wat.er Crud and 0.1 mci California F$1trate 255 3/16/72 DPR-6 G.E. North Carolina 3 U0 Fuel Rods cO.,1 mci 2
SNM-1097 256 3/30/72 DPR-6 NECO 16-NSF-1 (A-11) 5 Heat Exchangers 1.25 Ci Morehead, Ky.
257 4/6/72 DPR-6
-NPI, 19-12667-01 Irradiated Cobalt 542,000 Ci 258 4/18/72 DPR-6 NPI, 19-12667-01 Irradiated Cobalt 493,000 Ci ES>
29 om
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m P00R ORIGINAL
APPENDIX D (Contd)
LANE MICHICAN BIG ROCK SCRFENHOUSE:
~~
(G)lH FORMATION CENTER (REACTOR I I STACKh SUBSTATION I
4
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LEGEND (t,4/
s nU l BIG ROCK POINT E
l p
SITE MAP l
8 WITH ENVIRONMENTAL i
MONITORING STATION S
[i C""
@ -LETTERED STATION S -F ILM MONITOR ONLY g
I MONITOR
@ - NUMBERED STATIONS -FILM AND AIR l
31 BIG ROCK POINT NUCLEAR PLANT-DOSE ISOPLETHS t:
h 0.0/w?em O!RECTION NO 3
/
OFF-SHORE
/ MI
/
(All land trojeClory)
/
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/
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ALONG SHORE l
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DIRECTION NO. I residue )
.j e.i ON SHORE (All water trajectory)
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1 r4 P00R ORIGINAL 0
J'
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l NOTE: Arrovs' and numbers around periphen indicate direction from which s
vind is blowing.
I
I 32 BIG EOCK POINT NUCLEAR PLANT-DOSE ISOPLETHS SO MI f
~
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SCALE O
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APPENDIX D (Contd)
Sampling.and Analysis Summary Number of Samples' Frequency of Type Description' Location Collected Type of Analysis Analysis 31
. Air Continuous'~at
'All 179 Gross Beta, I
Weekly
.Approximately' i -
1 Cfm.
Lake Water 1 Gal Grab ST 12 Gross Beta, Gross Gamma Monthly 54 4"""t*"17 90Sr, 13hCs, 137c3, Mn, Co, Co, Zn, 59 5
Fe Well Water 1 Gal Grab ST 6
Gross Beta Monthly Gamma Dose Continuous
.All 163 Film Dose Monthly a
9 1
I l
l' l
APPENDIX D (Contd) l
'High,: Low and Average. Concentrations'
'For Highest Sampling Location l
-Type
' Type of Analysis Location High Low Average Air Gross Beta-Gamma PT
.o.69 pci/1 o.04pci/1 0.16pci/1
- I-131
.Bc o.27
<o.2 0.21 Lake Water Gross Beta BR ST LWO 11.8 4.9 8.8 Gross' Gamma BR ST LWO 13
<6 8.3 Sr-90 BR ST IMO 2.4 1.8 2.1
-Cs-137 BR ST.LWC 6
<2 4
Well Water Gross Beta BR ST W 54 2.1 12 Film Dose ST 19 Millirad 0 Millirad 12 Millirad Y
APPENDIX D (Contd)
Big Rocli Point Plant A Site Station Monthly Average Radioactivity e Average of Stations 2-5 Concentration in Air
+ Back6round Stations 6 and 7 Average lD0 90 40 70 t
60 50
.40 30 Q
.35 A
/
O!
30
/
.t5
/
2 A
/
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u ny
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.04 M
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- i. s r---
.Os*
JAN FEB MAR APR MAY JUN JUL AUG SEP OCT NOV DFC
t t e el lt C
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Appendix D (Contd)
Big Rock Samples May 1972 Nuclide (pCi/g)
Camma Beta 60 40 54Mn epa /g*
PCi/g 134Cs,58co 65Zn Co g
137Cs 95Zr Sample Shore Minnow Distharge 0.3 0.3 0.7 0.20 3.1 1 0.3 1/4 Mi.
E.
of Disch.
0.3 0.6 1.0 0.17 0.27 3.4 1 0,3 1/4 Mi.
W.
of Disch.
0.2 0.4 1.0 0.15 3.4 x 0.3 Nine Mile Point (a) 0.1 0.3 1.0 0.11 2.8 t 0.3 i
Mt. Mc Sauba (b) 0.2 0.8 0.06 1.9 1 0.2 Alcwife 0.1 0.1 0.8 0.09 2.6 1 0.3
. Lake Trout 0.5 1.0 0.13 3.1 1 0.3 OError is t 0.01 or 10%, whichever is greater (c) 3 miles East of discharge (b) 3 miles West of discharge d
Appendix D (Con *d)
BIG ROCK SAMPLES May 1972 Muclide (pCi/g)
Gamma Beta 95 134Cs,58co 65 n 60 Z
co 40g 54Mn epm /g*
pCi/g Sample 137Cs Zr Croyfish Discharge 0.4 0.1 0.3 0.36 2.2 2 0.2 Nine Mile Point (s) 0.2 0.3 7
0.20 1.6 2 0.2 Mt. McSauba(b) 0.3
?
0.13 4.0 2 0.4 Pefiphyton Discharge 1.0 0.8
?
0.9 7.7 104 2 10 1/4 Mi.
E.
of Disch.
1.2 0.3 2.2 61 ! 6 1/4 M1.
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 2 Filament Algae Discharge 0.18 51 2 5 1/4 Mi..W.
of Disch.
0.17 0.2 0.13 19 2
Nine Mile Point 0.1 0.17 1.2 0.19 13 2 1**
Mt. McSauba 0.3 0.8 0.17 17 ! 2 Bottom Sediment Disch. Midway 0.5 0.2 0.06 0.4 2.4 9.2 2 0.9 Disch. Off Shore End 0.5 0.2 0.05 0.4 2.0 16 ! 2 Disch. Shore Line 0.4 0.2 0.05 0.4 1.5 12 2 1 CError is 2 0.01 or 10%, whichever is greater 00 Abnormal amount of solids (c) 3 miles East of discharge (b) 3 miles West of discharge