ML20030A503

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Semiannual Operating Rept,Jul-Dec 1971
ML20030A503
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 02/28/1972
From: Sewell R
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
References
NUDOCS 8101090780
Download: ML20030A503 (36)
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Text

{{#Wiki_filter:, 75 Cf s . s vP *' CONSUMERS POWER COMPANY yr.S"* Docket No 50-155 Report of Operation of Big Rock Point Nuclear Plant Licence No DPR-6 July 1,1971 Through December 31,1W1 I. SUfMARY OF OPERATIONS A. Power Operation Due to premature failure of several fu'el bundles, the plant was operated at 53 MWe (gross) during this report period, to limit the heat flux on the fuel cladding. On July 15, 1971 during a routine test, the engine of the emer-gency diesel generator shut down automatically due to cooling water system high temperature. After diagnosis, the engine was restarted within several minutes and operated normally. Details of this failure to start were re-ported to the AEC by letter from Consumers Power Company dated August 11, 1971. A replacement cooling water pump has since been purchased and will be installed during the March 1972 refueling outage. The pump impeller was worn to' the extent that sufficient priming could not be completed with-out the aid of an additional priming water supply. The replacement pump has a larger head tank capacity than that of the original to allow for a greater priming capacity. The diesel vill be test operated upon completion of the installation. On September 1,1971, a settlement was reached between the union (Utility Workers Union of America) and the Company with the subsequent return to work by the members on that date (see Fourteenth Semiannual Report). The plant continued to operate at 53 MWe (gross) at an off-gas release rate of approximately 8,900 pCi/s until September 9,1971 when the plant was removed from service on a scheduled outage to perform the con-trol rod drive six-month testing program as required by the operating license. This marked the end of 97 days of continuous plant operation. 4 (M/0 fo 7To

g Additional work performed during the outage included: 1. The plugging of a main condenser tube leak (0.52 gp=) in the top row of tubes (north bank) and in the general area of previous failures. The tube began leaking just prior to shutdown and necessi-tated premature replacement of the condensate resins to maintain primary coolant conductivities within allowable limits. 2. Thereplacementof1/4-inchlubeoiltubingfromthediesel ge-erator engine to the lov lube oil pressure switch in the 1ccal control panel with 3/8-inch tubing was completed. The change was made in refer-ence to another plant which recorded trip-out problems in cold weather caused by poor oil transfer through small lines. The unit was returned to service at 53 ige (gross) on september 12, 1971aftera69-1/2-houroutage. Following the return to power, the No 1 recirculating pump inner seal cooling tempc:catures began rising, indicating failure of the inner seal. However, the seal cooling terperature remained within cperating limits. The pump has operated in this condition to date 'with the outer seal taking 90% of the reactor pressure drop. The plant continued to operate at 53 ige (gross) at an off-gas release rate of approximately 10,000 UCi/s until September 22, 1971 when the plant was manually scrcmmed following the loss of all of the plant's major rotating equipment including the main condenser circulating water pu=p s. The loss was attributed to an accidental tripping of the station power relays located inside the " Relay and Metering Potential Transfomer" cabinet mounted on the 2400-volt bus structure. The cabinets have since been labeled and tagged to prevent any reoccurrence. Control rod drive C-3 (previously stuck at notch 17) scra==ed satisfactorily; however, it could not be moved from the 00 position prior to start-up. It is planned to continue operation throughout the remainder of run 9 with C-3 at 00. The unit was returned to service on September 23, 1971 after an outage of approximately 18 hours. The plant continued to operate at 53 ige (gross) with the off-gas release rate of approximately 13,500 uCi/s until septe=ber 28, 1971 when the plant scra=med on high flux following loss of the 138 kV trans-mission line. The loss was attributed to an electrical stom which was

. a....., r.; q._..... -.. ;.y..,. e. g.,.3. _, ...y. ...7...,. .,.r. ...x ... -.,.;..., 7..;..... ,y,.,,. l 3 in the area at that time. Subseauent inspection of the plant revealed no abnormalities and the plant returned to nervice on September 28,19' 1 after an outage of approximately 14 hours. Upon start-up, the off-gas releaserateincreasedapproximately30%to21,500pCi/s,andtheNo2 reactor recirculation pump outer seal began to fail. The inner seal was taking essentially the entire reactor pressure drop. The seal leak off rate began to decrease; however, the seal temperatures remained within allowable limits. Failure of the seal on October 18, 1971 necessitated pump shutdown, causing the plant output to be reduced to 40 We (gross). Prior to reducing load, the off-gas activity was 32,800 pCi/s. After the load reduction, the off-gas rate dropped to 25,000 pCi/s. The plant continued to operate at 43 We (gross) until October 23, 1971 when the unit was removed from service to replace the No 2 reactor recirculating pu=p seal cartridge. Upon inspection, some fatiguing; of the seal materials was found as evidenced by the erosion and heat checking of the stationary face and rotating ring. A likely suspect is the improper assembly clearances which have since been cor-rected. This theory will not be proven until the life of the seal, which is now in service, is determined. A 30 se (gross) load rejection test was performed on October 24 to test the turbine bypass valve performance following adjustments made in the valve elettronic control system on October 23 The pressure in-crease in the reactor was 43 psi and the average out-of-core flux increase was 60%. This compares to figures of 50 psi and 50% for the 30 se (gross) test of April 1970. The unit was returned to a 43 se (gross) output en October 24 after an outage of approximately 33 hours. The off-gas release rate dropped to approximately 14,000 pCi/s. It was planned to limit plant output to 43 s e (gross) for the remainder of the run. This power level should mitigate fuel failures and maintcin off-gas limits at acceptable levels as vall as conserve reactivity to enable operation into March 1972. The plant continued to operate at 43 We (gross) with the off-gas release rate at approximately 21,600 vCi/s until November 26 when the g linkage arm of the turbine trip solenoid failed allowing a turbine and

4 generator trip. However, there was no scram since the turbine bypass valve dumped steam to the condenser upon sensing the increasing pressure. Repairs were effected and the plant returned to service at 43 MWe (gross) after an 11-hour outage. Off-gas release rates have increased gradually to 27,800 pCi/s to date. B. Notes of Significance The cause of the ASCO valve failures has been determined by the manufacturer and replacements will be made. (Please refer to the Addenda to the Twelfth Semiannual Report submitted to the AEC dated October 9, 1970.) when the valve is energized over a period of time, relaxation of the resilient 0-ring seat allows the plunger to creep. Thus, the important gap is lost which is necessary for the valve to begin movement. Fifteen metal-to-metal seat replacement valves will be in-stalled during the refueling outage. Statistical Operating Data For the Period July 1,1971 Through December 31, 1971 MWhe (Net) - 195,734.2 MWhe (Gross) - 206,578 - 641,763 MWht Hours Critical - 4,306.9 Hours Generating 4,270.1 Plant Availability Factor, % - 96.7 Plant Capacity Factor, % - 62.4*

  • Plant operated at a derate of between 22 and 32 MWe for preservation cf fuel integrity.

II. EFFLUENT AND ENVIRONMENTAL MONITORING A.. Introduction and Conclusions 1. Introduction Radioactive effluent data for the.first six months of 1971 were included in the semiannual report to the Commission covering the period January 1 through June 30, 1971. No data on environmental levels and radiation doses resulting from these releases were included in the g.

5 report, however, due to the fact that releases were well below the levels required to measure dose by sampling environmental media. Since that time, however, the Commission has published its safety guide Number 21, entitled " Measuring and Reporting of Effluents From Nuclear Power Plants," and is considering the publication of an additional safety guide on mea-suring and reporting of radioactivity in the environs of nuclear power plants. Because these guides require results of effluent and environmental monitoring programs to be submitted as part of each facility's semiannual report of operations, all of the 1971 environmental monitoring data obtained on the Big Rock Point program is included herein and correlated with all of the 1971 waste effluent data. Reports for 1972 and beyond will contain, on a semiannual basis, an effluent and environmental section for the appropriate six-month period. . Annual _ releases and doses can then be obtained by totaling the results of the two semiannual reports. Releases of radioactive material both to the atmosphere and Lake Michigan during 1JT1 were well within the facility-licensed limits and the Commission's regulations, particularly Title 10, Code of Federal Regulations, Part 20. Environmental levels of radioactivity as found in the vicinity of the plant were composed abnost entirely of naturally occurring radioactive materials and fallout'from nuclear weapons testing programs. Only in ths vicinity of the circulating water discharge canal was radioactive material of plant origin found. These materials occurred primarily in aquatic organisms. These levels of radioactive materials, however, were extremely low and pose no threat to the health and safety of the public. Further,'these levels of radioactive material found in the resident biological commanity are consistent with levels found in previous years and show no upward trend.- 2. Atmospheric Effluent Gaseous-releases to the atmosphere totaled 285,600 curies of fission and activation gases. This corresponds to 0 91% of_the licensed technical specificaticn limit of 1 Ci/s.. Particulate releases totaled 0.056 curie or o.15% of the licensed limit while halogen releases totaled ]. .-vr-

6 0.55 curie or 1.45% of the licensed limit. Tritium releases totaled 7.5 curies or 4.1 x 10 % of a Itmit based upon meteorological dispersion to the point of maximum ground concentration. The radiation dose at the point of maximum ground concentra-tion resulting from these releases was 1 9 millirem. The integrated dose to the population out to 60 miles was 7 3 man-Rems-3 Liquid Waste Releases Liquid waste releases totaled 3.46 curies of radioactive material. This release corresponds to 5 9% of technical specifications limits. Additionally 10 3 curies of tritium were released corresponding to 0.003% of 10 CFR 20 permissible concentrations in the discharge canal. Population doses based upon drinking water from the Charlcyoix municipal system was 0.002 man-Rem and consuming all of the Lake Michigan fish harvested was 0.41 man-Rem. B.' Atmospheric Effluent and Environmental Su= mary 1. Effluent Calculational-Methods A sample of off-gas is obtained weekly during power operation and analyzed by Cuma spectrometry for six noble gas radionuclides.* Based I upon the mixture of nuclides found, a stack release rate utilizing an appro-priate factor for the 30-minute hold-up period is determined. This release rate, usually expressed in tems of microcuries per second, includes a total of 22 noble gas radionuclides as rhown in Appendix A. Daily samples taken for the remainder of the week are analyzed on a gross basis with a similar decay factor applied.** - In reality, however, a pure recoil, equilibrium or diffusion mixture is seldom observed in the mixture sampled. Hence, decay factors appropriate to the mixture are assigned on a weekly basis determined by the gamma spectral results of the weekly sample. When a given weekly sample ia a combination of mixtures, an appropriate weighted decay factor is applied. s i . *The six nuclides are: Kr-85m, 87, -88 and Xe-133, -135, and -138. L

    • The decay factors for a 30-minute hold-up period are 0.43 for a recoil

-mixture, 0 59 for a diffusion mixture and 0.83 for an equilibrium mixture. r l l'

7 l 1 i i ' i Monthly totals expressed in this manner were shown in previous semiannual reports. In this and future reports, the individual nuclide releases will be shown, based upon the distribution measured at the off-gas monitoring point and applying the decay factor associated with each nuclide in the mixture. The turbine sealing steam system contributes less than 1% of the gas discharged and has the same distribution of nuclides as the off-gas. This is reflected in the monthly totals shown in Appendix A. Activation gas releases are composed primarily of N-13 The rate of release is power-level dependent and is incorporated in the total monthly releases shown in Appendix A. Particulate and halogen releases to the atmosphere are mea-4 sured by counting particulate and charcoal filters weekly. These filters collect stack effluent continuously at a rate of 3 cubic feet per minute. Determination of release rates in this manner assumes radioactivity is continaully 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. No gross alpha measurements on atmospheric effluent were made in 1971. Due to its low concentration in primary coolany, ($3 x 10-7 pci/ml) for 1971 compared to about 10-1 pci/mlforgrossbetaactivityandconser-vatively assuming alpha-emitting isotopes are as volatile as other particu-4 late nuclides released, the possible release of alpha-emitting nuclides is about 5 to 6 orders of magnitude lover than beta-emitting particulate releases. 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 humidity and 90% radiolytic gas by volume both at primary coolant con-centration to. determine tritium releases both in vapor and molecular form. b. Turbine Sealing Steam - The measured flow rate'at 100% relative humidity and primary coolant concentration. c. Containment Ventilation - The measured flow rate and mea- - sured' containment tailding concentration. ~ The results of these calculations'are also shown in Appendix A.. 4 t y f ,3 ,s -. + +,-y 4 '. w

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8 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 attempted then by measuring levels of radioactive materials in the plant's environmental surveillance program. For the Big Rock Point Plant, the average yearly meteorological relationship between release rate and downwind concentra-tion of radionuclides at distances from the plant up to 60 miles has been established. Using a typical equilibrium mixture of noble gas release 3 after a 30-minute decay, a concentration 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. FromthisandtheapplicableX/Qvalueof3.45x10 s/cm, the release rate required to produce a dose of 500 millirems per year at the point of maximum ground concentration is 2.4 curies per second. The licensed technical specification limit is 1 curie per second. Further-more, the licensed technical specification limit for particulates and halogens is (1.2 x 10 ) x (MPC) which produces a concentration at the _ point of maximum ground concentration per the chart some 2400 times below maximum permissible concentration allowed in Title 10 CFR 20 for I-131. Based on these values, the resulting doses can be calculated based upon 1971 release data as shown in i.ppendix A. These doses are then corre-lated to measured environmental levels if the releases are sufficiently great to observe in environmental media. The environmental surveillance program includes continuous sampling of air for particulate and halogen activity at ceven 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 determine 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 environment. Any increase in direct-dose at the closer stations can then be attributed to be of plant origin due to direct radiation from the plume.

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9 L 3 Environmental Dose Calculations The dose at the point of maximum ground concentration and total integrated dose in man Rem for 1971 is shown on the following page utilizing the following facts and conservative assumptions: A 2.4 Ci/s continuous release produces a dose of 500 mrem / a. yr at the point of maximum ground concentration. b. The total astimated 1971 population resides 24 hours per day all year in the same location. The 1971 population is estimated from 1960 population numbers available in the FHSR corrected on a county by county basis utilizing 1970 census figures. c. No shelter credit is taken which significantly reduces the dose due to the fact that over 50% of the energy emitted by the noble gas mixture is in the form of beta particles. An equilibrium mixture was used to calculate doses as this mixture most closely represented 1971

releases, d.

The calculated doses then are primarily skin doses due to the large beta fraction of decay energy. Hence, the dose rate and integrated doses shown are not whole body doses sud the term " man-Rem" is somewhat con-servatively inappropriate. e. Semi-infinite cloud geometry wes utilized to calculate doses. f. No correction credit was taken for radioactive decay due to the time required for the p1tme to travel 60 miles. Doses from particulate, iodine and tritium releases as shown in Appendix A were negligible compared to that received from noble gases due to the conservative limits in the plant technical specifications and the absence-of the milk food chain in the area affected by the plant. In order to obtain greater sensitivity of measurement, a com-parative program of-film vs thermoluminescent dosimeters (TLD) was started in late ICJ71. The program consists of placing a film and TLD side by side -at'each monitoring station for a one-month. exposure' period. Comparative results will be available by mid-1972. If better exposure sensitivity and consistency are obtained'through the use of TLD, the environmental sur- ~veillance program will include use of these dosimeters in the future. ~.(-

CALCULATED RADIATION DOSES FBOM GASEOUS EFFLUE!ITS,1971 Avg Pa = pose Annulus Avg Distance X/Q 1971 Population (Pa)(X/Qn)(G)** Rate Miles Miles Meters (s/m)3 X/Q,n _ Pa (Pa X/Qn) Man-Rem /Yr mrem /Yr -8 1 1-2 15 2,420 3.4 x 10 (1) 86 8.6 x 10 0.16 1 9*** 2-3 2.5 4,03o 2 5 x 10-8 (.80) 515 4.1 x 10 g,77 2 3-4 35 5,640 1.4 x 10-8 (.40) 1,080 4.3 x 10 0.81 2 3 4-5 4.5 7,250 9.o x 10-9 (.26) 4,400 1.1 x 10 2.w 5 - 10 7.5 12,080 3.7 x 10-9 (.u) 4,940 5.1 x 10 o.96 2 -9 2 lo - 20 15 24,160 1.1 x 10 (.03) 20,000 6.0 x 10 1.13 20 - 40 30 43,400 4.2 x 10'10 (.01) 26,100 3 2 x 10 0.60 2 2 40 - 60 50 80,500 1 5 x 10-10 (.004) 100,900 4.o x 10 g,75 {Pa=158,037 [Da = 7 25

  • Normalized X/Q to maximum X/Q.

Q Curies /Yr Released 2E,h

    • G = Q Curies To Produce 500 mrem /Yr x 0.5 Rem = 75,666,400 x g,5,g,gg19
      • Point of maximum ground concentration.

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11 2 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 conditions, the following maximum concentrations can be calculated: ~1 3 Particulates (October) (1.2pCi/s)x(0.005)x(3.4x10 s/cm) 2.04 x 10-16 pCi/cm3 = Halogens (September) (1.2pCi/s)x(0.015)x(3.4x10~1 s/cm) 3 6.1 x 10-16 Ci/cm3 (' = f These compare to the minimum detectable activity values and normal background concentrations as follows: Maximum Calculated Minimum Detectable Normal Background Release Concentration pCi/cm3 Activity pCi/cm3 Acti7ity UCi/cm3 Particulate 2.0 x 10-16 1 x 10-1" 7 x 10~1" Halogen 6.1 x.'10-16 1 x 10-12 Hence, the negative data obtained in the program was expected. C. . Liquid Effluent 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 flow rate of 48,000-52,000 gpm dilution for liquid effluents is obtained i through the use of the condenser circulating water pumps, two at 24,000 - gpm each, and house service water pumps, two at 2,000 gPm each. Each collected tank of liquid is sampled, analyzed for radio-l - active-content, and discharged at a controlled rate to assure that pcr- !~ nissible aoncentrations are not exceeded in the-canal prior to dilution in Lake Michigan during the time of discharge. Each sample is analyzed by gamma. spectrometry to. identify as many of the cceponent nuclides as .possible.. (See Appendix B for results.) Pemissible concentrations in the canal are detemined from the following: i \\. / y e -y e .a v,- g

12 I Ciiiy MPC where Ci is the concentration of the i_th isotope in the canal at the given concentration measured in the tank diluted by the known canal flow rate. Those isotopes not identified by ga=a spectrometry but mea-sured by a gross beta-ga=a analysis are presumed to be Sr-90 and re-leased on that basis. This is a very conservative assumption since, in reality, only a small fraction of the unidentified portion is strontium and, of that fraction, the majority ia probably Sr-89 whose MPC is 100 y times larger than that of Sr-90. Measurements will be made to confirm these fractions and the results will be applied to future reports. Gross alpha activity has not been measured on each sample due to its very low concentration in primary coolant. Tritium measurements are not made on each batch released. Primary coolant tritium concentration averaged 3 x 10-2 uci/mlduring 1971 and 60% of liquid waste volume nomally is primary coolant. The other Wp is made up of either laundry or condensate demineralizer regen-eration rinses which contain no tritium. Hence, the tritium released is calculated on the basis that the average concentration in each month is 60% of primary coolant concentration. During April 1Cf/1, a plant.7perating procedure was changed which significantly reduced the liquid effluent volume. This change in-cluded the elimination of all condensate demineralizer regenerations with the resulting spent resins presently being stored on site for future dis-posal ac solid waste. Elimination of the regeneration cycle resulted in a significant reduction in liquid effluent and radioactivity release (see Appendix B). Furthermore, it allowed the recycling of greater amounts of collected water back into the primary coolant system. This further re-sulted in only about 10% of the liquid effluent originating from the primary system. Therefore, for the months of May-December 1CJ/1, tritium releases are calculated' on the basis that 10%.of. the released effluent originated as primary coolant. 2. Envircamental Methods and Data In order to predict potential radiation doses resulting from .(. the above releases, environmental transport and uptake factors must be

4 13 i l known. A confimaticn of these calculated doses is then atte=pted by measuring levels of radioactive caterials in the plant's environmental radiation surveillance program. At the Big Rock Point Plant, daily ccm-posite condenser circulating water inlet and canal water discharge sa=ples are taken and analyzed for radioactive content. In addition, a monthly co=posite of these samples is analyzed for radioactive content. These results are shown in Appendix D. Additional aquatic sa=ples are taken and analyzed during the sumer Growing season and these results are also tabulated in Appendix D. Based upon the release of 3.459 curies of radioactive material (less tritium) which results in an annual average concentration in the dis-charge canal'of 3.4 x 10-8 pCi/ml,theanalysisofdischargecanalwater should indicate an increase of radicactive material in discharge canal water samples since the minimum detectable activity for gross beta mea-surements is about 5 x 10-9 pCi/mlorabout7timeslowerthantheaverage concentration discharged.. The results shown plotted in Appendix C indi-cate an average of about 3 0 x lo pCi/ml for the year, in close agree-ment with the predicted 3.4 x 10-8 9C1/=1. The nearest municipal drinking water supply intake is located-in Charlevoix, Michigan, which is generally. upstream of the prevailing current flow in Lake Michigan at this location. Mcwever, since current patterns do occur that could, at times, carry the discharged water in the direction of Charlevoix, population dose based upon this flow all year is . 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. Other water supplies exist in Little Traverse Bay but due to their relatively r o ote locations in. relation to.the plant the contribution to total population dose is not significant. Additionally, radiation dose to human populations can occur - as a result of plant releases through the consumption of -fish residing in ~ .the area of the plant discharge assimilating a portion of radioactive ma-terial discharged. The aquatic sample analyses shown in Appendix D were used for these calculations as shown in'the next section of the report.

14 3 Environmental Dose Calc tlations Utilizirg the measured values of radionuclides released as shown in Appendix B, the following ior=ula, and the standard man model, drinking water doses can be calculated as follows: D, = Cii ' (Limiting Dose Rem /Yr) I L ij where: D,istheindividualdoseinRem/yr, Ci_ is the yearly average concentration of the individual nuclidesmeasured,inpCi/ml, MPC is the concentration of each nuclide measured required to produce the limiting dose at continuous intake in pCi/ml and Limiting Dose is the dose produced at continuous exposure to MPC concentrations. In calculating ingestion dose from the consumption of fish, a similar equation is used except that a standard daily diet of 50 grams of fish flesh is used in contrast to the 2200 ml of water consuned daily by the standard man. This, in effect, alters the MPCi by 50/2200or0.0227 The aquatic biota and fish analysis data are shown in Appendix C. The analytical results of fish samples were obtained only on the one sa=ple collected offshore at Big Rock Point in October of 1971. Two sa=ples will be collected in 1972 to provide a more representative average. The calculation of individual doses, both from drinking water and consuming fish, are per the previous formula while integrated popula-tion doses in man-Rem are calculated utilizing the following parameters: a. For drinking water, the individual doses are su=med over the population of Charlevoix, about 2500 people. l b. Forfishconsumption,anaveragevalueof1/800ofthe . concentrations measured is used as an average for all Lake Michigan fish. This is conservative due to the fact that a dilution factor of at least 800 is reached by the time the discharged water is only 3-5 miles from l the plant. - These concentrations in turn will be used for all fish har-vested during 1971 for Lake Michigan. The data on fish harvest for Lake Michigan is available from the Michigan Department of Natural Resources and appears in Appendix C. i I l

15 f. Both of the dose calculations are conservative in that: a. Equilibrium is not attained in the human body particularly for the unidentified portion of the release assumed to be Sr-90. b. No credit is taken for precipitation and deposit in sedi-ment on uptake by life foms other than fish which is seen to occur by the data shown in Appendix D. Additionally, the drinking water dose calculation assumes the prevailing current flow in the direction of Charlevoix continuourly while the fish consumption calculation assumes all radioactivity foundi except K 10, to be due to releases of plant ori in. Inreality,mostIEthe 6 Cs-137 is in all likelihood due to fallout. e e i i a i I. (. L_

.= ,.s CALCULATED RADIATION DOSES FROM LIQUID EFFLUENTS, 1971 Concentration Ci/MPCi [Ci/MPCi MPC Critical Released Whole Body Thyroid Bone -Whole Vzctor Isotope pCi/ml Organ Avg uCi/ml (0.5 Rem /Yr) (0.5 Rem /Yr) _5 Rem /Yr) Body Thyroid Bone ( Wat'er Zn 65 .1x10-Whole Body 6.3 x 10-9 6.3 x 10-5 ~ -3 Cs-134- '9x10-6 Whole Body 1.0 x 10 1.2 x 10 .H-3' 3x10-3 Whole Body 1.0 x 10-7 3 3 x 10-5 1 3x10-3 .I-131 3x10 Tm roid 8.5 x 10-9 2.8 x 10-2 2 9x10-2 -2 -2 Other(l) 3x10-7 Bone 8.7 x 10-9 3.0x10 2 9 x 10 Fich 'Cs-137, 9x10-4 Whole Body 1.0 x 10-6 1.1 x 10-3 pCi/g Zn 65' 4.hx10-3 Whole Body 0.1 x 10-6 2.3 x 10 -5 1.1x10-3 UCi/g Doses: A. Drinking Charlevoix Water mrem /Yr 0.00081 Whole Body; O.018 Thyroid; 0.19 Bone B. Population Drinking Water Dose Man-Rem 0.002 Whole Body; 0.045 Thyroid; 0.47 Bone C. Eating Fish mrem /Yr 0.55 Whole Body D. Population Fish Consumption Dose (2) Man-Rem 0.41 Whole Body (1) Assumed to be Sr-90. ( l Using 23,873,689 pounds of fish harvested from Lake Michigan in 1970. This number includes both co:::mercial and sports catches as shown in Appendix C minus alewives which are generally not consumed.

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m l t l AfPENDIY B PP1T9T or RADIOACTIVE 1.1 QUID FIfWm? Pfl#.ASES I feettity: Big Pod Peint Wueleer Plant Docket! $0-155 Tear g i I tette Jon feb mr Apr my June ' July Aug feet oct

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  • '"I '* * #

I' 3. Percent of appiteoble Statt = " " *8* tmery err lent for the monthe January through Aprtl and 104 the rest of the yeer Desed upon the feet that shout los of relenee4 Itqu14e is enndensete or {' uct/mi for 1771. 4 end that the tritime concentretton in primary coolent everagel 5 s lo* l 5. 1W nuclideo Co.134 end 137 both oppear terularly en the gassen spectret enelyste of coch botch. Their gamme decay energies are sistler and therefore difficult t to separete free one another. Since C -134 has the more restr9etten eGv", perateelble release calculottome seeiamo all costum activity to be due to Os.134 6 All remaining setty!ty, te, 3.b5%?.573 = o.MlV, C1 1e eesume4 to be br.% and t o releised unMr this restrtett re MPC. "co

w APPENDIX C .'0FF-SITE SHIIMEITIS OF RADIOACTIVE MATERIAL A'totalc of 28 shipments ofl radioactive material off-site were made during 1971. Details are as folluss: Shipment Transfer Radioactive Material No Date'- From Transfer To 221 1/22/71 .DPR-6 Babcock and Wilcox,' SNM-778, Lynchburg, Va In-Core Wire 50 Ci 222. 1/26/71 'DPR-6 GE-Val, 0017-60, California Feed-Water Crud & Filtrate 0.2 nci 223 1/27/71-DPR-6' NECO 16-NSF-1(A-11), Morehead, Ky 2400 Gal of Low-Level Waste 1.0 Ci 224'- 1/29/71; . DPR NECO 16-NSF-1(A-11), Morehead, Ky 2400 Gal of Low-Level Wste 1.0 Ci .225 -3/1/71. 'DPR-6 GE-Val, 0017-60, California Fuel Inspection Tools 19 nci -226 3/17/71 DPR-6 NPI 19-12667-01, DOT SP 5364 Irradiated Cobalt 479,230 Ci

227 3/23/71 DPR-6

'NPI 19-12667-01, DOT SP 5364 Irradiated Cobalt 507,360 Ci 228 3/24/71 DPR-6 GE-Val, 0017-60, DOT SP 5971 Dummy Fuel Rods 100 Ci 229 3/24/71 DPR-6 GE-Val, 0017-60, DOT SP 5607 Centermelt Fuel Rod 1,000 Ci 230-3/26/71 -DPR-6~ NPI 19-12667-01, DOT SP 5364 Irradiated Cobalt 267,600 Ci 231 3/29/71 DPR-6 NECO 16-NSF-1(A-11), Morehead, Ky 115 55-Gal Barrels Iow-Level Waate 03Ci -232 4/9/71 DPR-6 Byron-Jackson Pump, Cri43-74, Los Recirculating Pump Parta 0.2 mci Angeles 233 4/14/71 'DPR-6 GE-Val, 0017-60, California Feed-Water Crud & Filtrate 0.2 mci 234 4/27/71 .DPR-6 GE-Val, 0017-60, California Fuel Inspection Tools 0.2 mci 235 5/28/71 DPR-6 GE-Val, 0017-60, Californin Feed-Water Crud & Filtrate 0.2 nci e

u -APPENDIZ C (Contd) Shipment. Transfer No' Date -From Transfer To Radioactive Material 236_ '6/7/71 DPR-6.- Palisades Nuclear Plant, 21-08606-05, Michigan ~ Steam Drum Crud Sar:ples 3 mci 4 I 237- '7/.6/71 DPR-6 'NFS, CFS-1, New York 10 Spent Fuel Assemblies 2,478,520 238 7/6/71-DPR-6 New Jersey Nuclear, Richland, Wash Crud Chip-Fuel Bundle 6 mci f 239 7/9/71 DPR-6 GE-Val, ' 0017-60, California Feed-Water Crud & Filtrate 0.1 nCi ~240 -7/9/71-DPR-6 GE-Val, 0017-60, California Feed-Water Test Filter 0.5 nCi j 241 7/14/71 DFR-6'. GE-Val, SNM-960, DOT 5607 - 4 Spent Fuel Rods 17,000 Ci .242 7/23/71 DPR-6 'NFS, CFS-1, New York 10 Spent Fuel Assemblies 2,035,171

243 8/11/71

_DPR-6 N78, CFS-1, New York 10 Spent Fuel Assemblies 2,466,140 2hh 8/30/71 DPR-6 ' NFS,. CFS-l', New York 9 Spent Fuel Assemblies 2,094,734 l l 245 9/15/71- .DPR-6 NFS, CFS-1, New York 6 Spent Fuel Assemblies 1,495,162 246' 9/27/71 DPR-6 GE-Val, SNM-960, DOT SP 5607 5 Spent Fuel Rods 10,000 Ci 247 12/28/71

DPR-6.-

Palisades Nuclear Plant, 21-08606-05 Calibration Sources 0.1 mci - 2h8 12/29/71 DPR-6 SWI 9-775, San Antonio, Texas Ultrasonic Probe 0.1 nci E$

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I y R O T Y I L N N O O M R R O I T A T L S IN D A N N T C N O NI M A I O P_ E T D P A MA M M T N M N L L S O I I E K F F R G C E I G VN E O T I S S L R I N R N N EO O O S T I T HI T I G T N A A O T T I I B WM S S D O E E R RE E B TT M E U Il L N Y__ h@ g / ) dt J /_ II g no C ( EV 6 D N AN %a) C X A O I D I o T D C A A N E M R y E t E EL P H SV R O P l l8 UAE O T D F CN A C OC I N AO b S l E T H D I T@(RA M NA U T E E E L S B R@ c U E S S iI ! K A K h f oR* L C O R K C G A I T B S ,,15 geanpc I' i 33 hNE 1 @3

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APPENDIX D (Contd) Big Rock Sampics _ October 1971 Nue]ide (pCi/g) Gamma Beta Srmple Cs Zr Cs,58co 2n_ Co K ,cym/g PCi/g 65 60 40 Algac-Discharge 1.3 ? 0.2 < 0.1 0.70 2 0.07 ,.3 ! 0.1 Algae-k mi. E. of Disch 0.8 < 0.1 < 0.1 0.6 0.53 1 0.05 .5 ! 0.1 Algae-k ni. W. of Disch. 0.9 0.1 < 0.1 0.6 0.56 ! 0.06 7.8 ! 0.1 Algae-Nine Mile Point 1.2 1.0 0.53 2 0.05 11.1 2 0.2 Algae-Mt. McSauba 0.7 0.7 0.35 0.04 11.4 2 0.3 Pariphyton-Discharge 2.0 0.8 0.3 0.5 5.94 0.6 50.5 2.3 P2riphyton-k mi. E of Disch. 1.8 0.8 0.3 0.5 2.66 0.26 36.0 1 2.0 Periphyton-( mi. W. of Disch. 0.9 0.4 0.3 ? 0.79 1 0.08 39.0 2 2.3 Piriphyton-Nine Fule Point 0.3 0.4 ? 0.4 0.92 0.09 25.0 ! 4.1 Pariphyton-Mt. McSauba 0.5 0.3 ? 0.2 ? 0.49 1 0.05 28.6 ! 2.0 Crcyfish-Discharge 2.1 0.4' 1.7 0.2 ? 1.20 1 0.12 0.53 1 0.04 Crayfish-k mi. E. of Disch. 1.7 0.5 2.1 ? ? 1.21 ! 0.12 2.65 ! 0.20 Crcyfish-4 mi. W. of Disch. 1.0 0.3 0.7 0.2 12.6 i 1.3 1.26 ! 0.08 Crayfish-Nine Mile Point 0.3 <0.1 0.3 < 0.1 0.31 ! 0.03 1.51 ! 0.08 Crcyfish-Mt. McSauba 0.2 0.4 0.2 0.8 ? 1.36 1 0.14 0.76 0.06 Minnous-Discharge 0.6 0.2 1.1 0.49 1 0.05 2.38 1 0.07 Minnovs-Nine Mile Point <0.01 0.68 2 0.04 Minnows-Mt. McSauba ? 0.13 1 0.013 0.26 1 0.05 . Lake Trout . 0. i6 1.1 0.19 1 0.02 5.9 0.3 Parch 1.3 1.2 0.44 2 0.04 7.5 2 0.9 B; s 1.1. ? 0.1 3.1 0.27 2 0.01 3.2 2 0.1 (3

APPENDIX D (Contd) Big Rock Sac:ples June 1971 Nuclide(pCi/g) 95 Sample Cs 2r 'Cs, Co 65, 60 40 M 7 Co g Hn p g Cf g Pariphy ton-Discha rge 3.4 1.3 0.9 1.5 6.96 2 0.70 57.6 2.3 Pariphyton-\\ mi. W. of Disch. 1.1 0.3 0.1 0.2 1.3 1 0.13 43.6 2.3 'Pariphyton-\\ mi. E. of Disch. 1.3 0.6 0.3 0.6 4.89 1 0.49 89.312.6 Piriphyton-Nine Mile Point 0.3 0.2 ? < 0.1 0.9 0.38 1 0.038 23.2 1 2.0 P;riphyton-Mt. McSauba 0.6 0.2 0.2 < 0.1 0.9 0.66 2 0.066 32.1 ! 2.1 Diacharge Sediment.(1) 0.5 0.3 < 0.1 < 0.1 0.2 1.13 ! 0.113 16.3 1 2.0 Diccharge Sediment (2) 1.7 0.5 0.4 < 0.1 0.5 2.54 2 0.26 25.8 1 2.0 'Diccharge Sediment (3) 2.1 O.2 0.2 < 0.1 0.8 6.2310.62 17.6 1 2.G Algae-k mi. ' E. of Disch. 0.5 0.4 0.2 0.4 1.71 0.171 21.6 0.3 Algae-Mt. McSauba 0.1 0.5 0.43 t 0.043 19.0 ! 0.6 Pil. Algae-Discharge 0.3 0.4 0.9 0.2 1.52 1 0.152 16.6 0.7 Algae-\\ mi. 0, of Disch. 0.2 0.2 ? ? 0.29 2 0.029 5.2 0.4 Algae-Nine Mile Point 0.7 0.40 -0.04 8.5 2 1.1 Crcyfish-Discharge 1.3 0.2 1.3 0.3 2.0 2 0.20 1.49 1 0.05 Crtyfish-Mt. McSauba 0.3 0.4 0.30 1 0.03 1.10 1 0.06 .Crcyfish-\\ mi. W. of Disch. 0.6 0.8 0.43 1 0.043 1,56 1 0.08 Crcyfish-\\ mi. E. of Disch. 1.2 0.2 1.6 0.2 1.17 2 0.117 3.31 ! 0.14 Crcyfish-Nine Mile Point 0.2 0.1 ? 0.21 2 0.021 2.55 2 0.16 Shtre Minnows-k mi. W. of Disch. 0.2 0.5 0.18 ! 0.018 3.88 0.11 Shsre Minnows-( mi. E. of Disch. 0.2 0.4 0.122 0.012 2.14 1 0.08 Shsre Minnows-Nine Mile Point 0.2 0.8 0.1/ 2 0.017 1.59 1 0.08 Shore Minnows-Discharge 0.5 1.8 ? 0.63 0.063 1.10 2 0.05 Sherc !!!nner : ".t. Mesauba 0.2 0.1 0.12 ! 0.012 3.31 1 0.10 E-l

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28 d APPENDIX D (Contd)

  • Michigan Department of Natural Resources Fisheries Division Lansing, Michigan g

s Sports Catch, Lake Michigan and Anadromous Streams, 197_0 Species Number Caught Estimated Total Weight Perch 1,700,000 283,333 Walleye 69,000 207,000 Bass-246,000 492,000 Panfish 1,300,000 260,000 Northern Pike 146,000 292,000 Suckers 482,000 1,446,000 Smelt 2,800,000 280,000 Lake Trout 2h5,000 1,715,000 Rainbow Trout 285,000 1,h25,000 . Brown Trout 168,000 '840,000 Brook Trout 125,000 250,000 Coho Salmon 534,000 5,340,000 Chinook Salmon 180,000 2,700,000 -t Other Species 368,000 368,000 Total -- 8,648,000 15,898,333 i

  • Unpublished'1970' data from postcard' census program-of the Ndchigan Department of l

Natural Resources, Fisheries Division. s l

29 APPENDIX D (Contd)

  • Nichigan Department of Natural Resources Fisheries Division Lansinic Michigan 3

C000ERCIAL CATCH, LAKE MICKIGAN - - 1970 Species 1970 Alewives 5,981,415 Bu11 heads 610 Burbet 51,261 Carp 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 Wa11 eyes 7 White Bass 1 Whitefish 1,417,834 13,956,771

  • Less than 50 pounds.

Taken from: GREAT IAKES FISHERIES 1970 data taken from December issue of Michigan, Ohio and Wisconsin Landings. i

30 III. PRINCIPAL MAINTENANCE PERFORSED A. Heating Boiler The plant heating boiler was shut down for annual inspection. The handhole plugs and manhole were removed and the water side was washed and cleaned. All flue tubes were wire brushed; also inspected were the high-and low-level switches, the oil burner and firebox. A new water glass was installed. B. Plant Cranes Maintenance on several of the facility cranes was performed during the quarter. Several worn relay contacts were replaced in the con-trol circuitry of the auxiliary hoist portion of the turbine building crane. The reactor building gantry crane had all the switch contacts and relay con-tacts cleaned on the 75-ton vinch portion. The work was required due to sluggish operation in the slow-speed mode. In conjunction with recommen-dations made by the Safety Audit Review Board, the fuel handling winch over the spent fuel pool was given a complete overhaul including the in-sta11ation of a mechanical stop mechanism to act as a backup to the elec-trical upper limit switch. A Whiting Corporation crane inspector made a ccuplete inspection including electrical control of the 75-ton reactor semi-gantry crane and the 75-ton fuel cask loading hoist. Minor maintenance was performed as a result of the findings. C.- Control Rod Drive Pumps The No i rod drive pump packing was replaced' due to water spraying from the center piston. The replaced packing had been in service for 22 months and 10 days - a record run for this pump. A crankcase oil leak and water leak from the compression chamber were repaired by instaning new inner and outer oil seals and a new cylinder head gasket. The No 2 rod drive pump No 1 piston packing gland-sleeve and packing gland nut were replaced due to cross threading of the gland nut. New piston packing was also installed. D.1 Containment Sphere-Maintenance of 'the' insulating material on the reactor containment sphere was again required this summer. The one half inch of Insul-Mastic

i. 31 l is subject to cracking in the areas exposed directly to the afternoon sun. This inL turn allows water to penetrate the insulation and deterio - rate the bond between it and the steel structure of the sphere. These areas were repatched, sealed and painted. E. Reactor Feed Pumos The lubrication has been changed in No 1 and No 2 reactor feed pump couplings from 150 weight oil to No 1 Texaco Marfax grease. This was done to alleviate the problem of oil drippings from the coupling between the integral end rings and the external boss. F. Hypochlorite Tank The hypochlorite tank was drained, cleaned and inspected. The internals of the tank are in good condition. The tank outlet pipe to the connecting fle.nge had developed a leak. Repairs were effected by in-stalling a new section of PVC pipe inside the outlet nipple of the tank and the use of fiber glass cloth and resin at the flange connection. G. No 1 Stack Fan Excessive vibration and noise on the No 1 stack fan were elimi-nated by installing a new key and two additional Allen setscrews on the motor drive pulley.. The pulley, keyway and shaft showed considerable wear as evidenced by the loose fit. A new motor drive pulley is on order and will be installed'on arrival. H. No 2 Recire Pump Seal Replacement The No 2 recire pump seal which had failed in October was re-moved and inspected under the supervision of a Byron-Jackson field engi-neer. During the uncoupling procedure, the shaft travel was measured; it was within the proper specifications but at the high limit (0.005 'to 0.008 inch). The pump drive shaft coupling key, when removed, exhibited rub marks on the shaft sleeve end. The excessive key length explained the extraordinary tight fit when coupling the shaft together in the past. One sixteenth of an inch of material was removed from the key end. Excessive crud and foreign material was found deposited on the internal components of the seal assembly. The inner seal was found in. good condition with the exception of a chip broken out of the inner radius of the stationary carbon face. Since the chip was found and in one piece,

32 f nr disassembly of the it was thought to be broken off during mov + seal. The outer seal exhibited considerab. 'th a definite flow path about one inch wide across the statiot

n face. Tiny cracks across the rotating face ring could not be identified but were thought to be heat checks. No explanation could be given since the inner seal showed no such marks and seal temperatures remained well within their limits (115 actual with a maximum temperature limit of 180 F). Both U-cup assemblies and the cartridge shaft sleeve exhibited little wear and appeared to be in good condition.

It was theoriced that improper assembly clearances may have caused excessive pressures on the seal assembly. These have since been corrected and should aid in a longer seal life. I. Plant Discharge Canal Following the recent dredging of the condenser cooling water discharge canal, the east side of the canal as well as the northern tip of the _ west side have been observed to be eroding away very rapidly. The cause is attributed to a higher-than-ordinary lake level along with the annual. shifting of direction of the prevailing storms. The autunn northerly winds blow across the canal. Approximately 60 cubic yards of . concrete aggregate have been placed in an atte=pt to control erosion. J. Turbine Generator Exciter An exciter brush arcing problem occurred in November. Obser-vation under a strobe light revealed that the collector ring had accu:m-1. lated deposits of an oil-or grease-like substance. This substance appeared to have been sprayed onto the collector ring through the ven-tilation system. Once on the collector ring, the substance had been picked up by the brushes which in turn glazed the collector ring surface with a carbon-substance (carbon-oil) film.- This film was causing an extreme arcing problem. - Several different methods of cleaning the col-lector ring were attempted and subsequently rejected as ineffective. The' brushes were then individually removed and the contact-areas scraped clean. This reduced the film deposit on the collector ring significantly and eliminated the arcing problem. This procedure will be repeated and } - cleaning of the_ exciter and ventilation system will be initiated in an effort to correct the arcing problem. 4 v v + w

9 33 ( K. Continuous Air Monitor The new CAM has been installed on the 24-inch line from the exhaust plenum for primary coolant small-break leak detection in the re-circulating pump room. The CAM is in operation and has been calibrated with simulated I-131. Radiciodine will be ordered for a more complete calibration of the CAM. The history for this installation was reported to the AEC by letter from Consumers Power Company dated September 29, 1971. IV. CHANGES, TESTS AND EXPERIMENTS PERFOIMED PURSUANT TO 10 CFR 50.59(a) A. Facility Changes Reactor Enclosure Pipeway Dew Cell - Facility Change C-158 was completed on the pipeway dev cell. This change consisted of relocating the dev cell from its location in the pipeway return air duct to the pipe-way exhaust duct (just prior to entry in the exhaust plenum). This change was made to provide greater coverage.for any steam leaks developing in the pipeway area, especially those in the lower elevations, as all air leaving the pipevay will be monitored. Emergency Diesel Generator - Facility Change C-159 was ccepleted on the diesel' generator control scheme. This change consisted of in-stalling an undervoltage alam on the 24-volt, d-c battery used for crank-ing and control purposes (set point for the alam is 24.0 + 0 3 V d-c with a time delay of approximately 40 seconds). Reactor Enclosure Clean and Dirty Sumps - Facility Change C-162 was completed on the clean and dirty sump pumps. This change consisted of the installation of running time meters for each sump pump. i The' meters have been installed in the control room main control j panel (Panel CO-2, Section 4) and arr capable of reading 10,000 minutes in l 0.1-minute increment s. Four additional meters are also mounted on the meter panel and - are provided for future connection to the turbine building sump, the off-gas drain pump and the radwaste sump pump. B. - Modifications Post-Incident and Core Spray System - Modifications were per-formed to the redundant core spray valve electrical schemes for MO-7070 { and MO-7071 to eliminate the 40-second time delays associated with opening i + L

34 \\ and reactor pressure sensors in each scheme (IGllE, -F, -G, -H) were re-calibrated to operate at 200 psi (previously 150 psi). Assistance was provided in checkout of the schemes following modifications. The following flow alaras were calibrated and placed in service: FS-2520 Reactor Containment Spray System Flow Nomal - 100 gpm FS-2521 Reactor Containment Backup Spray System Flow Hormal - 100 gpm FS-2522 Reactor Core Spray System Flow High - 500 gpm FS-2523 Reactor Core Backup Spray System Flow High - 500 gpm C. Tests Component Leak Rate Test - The semiannual cor.:ponent leak rate test was performed in October 15T/1. The results of the test showed a component leakage rate of 41% of the mwimm allowable leakage rate as defined in the Technical Specifications. ~ Shutdown Margin Test - A shutdown margin test was performed in September 1971. This test demonstrated that we could have any two control rods out of the core and still meet our shutdown requirements. V. PERIODIC TESTING PERFORMED AS REQUIRED BY THE TECHNICAL SPECIFICATIONS The following tabulation shows the required frequency of testing, s plus the testing date-of the systems or functions, which may be periodically tested per Technical Specifications: -System or Function Frequency of Dates Undergoing Test Routine Tests Tested l - Control Rod Drivce Continuous withdrawal and inser-Each major refueling and 9/11/71 tion of each drive over its at least once every six i stroke _with normal hydraulic months during periods of 51 e e ds. 'a I Withdrawal of each drive, stop-Eachmajorrefuelingand 9/11/71 . ping at each locking position to at least once every six check'latchingiand unlatching months during periods of - _ operations and the functioning power operation.- of the position indicstion system. (

35 i System or Function Frequency of Dates Undergoing Test Routine Tests Tested Control Rod Drives (Contd) Scram of each drive from the Each major refueling and 9/10/71 fully withdrawn position. Maxi-at least once every six mum 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 its Each major refueling but 3/3/71 entire stroke with reduced hy-not less than once a draulic system pressure to de-year. termine that drive friction is normal. Control Rod Interlocks Rod withdrawal blocked when any Each major refueling but 2/26/71and two accumulators are at a pres-not less frequently than 3/3/71 sure below "f00 psig. ,once every twelve months. Rod withdrawal-blocked when two Each major refueling but 3/3/71 of three power range channels not less frequently than read below % on 0-125% scales once every twelve montis. (or below 2% on their 0 40% scales) when reactor power is above the minimum operating range of these channels. Rod withdrawal blocked when Each major refueling but 3/3/71 . scram dump tank is bypassed. not less than once every twelve months. Rod withdrawal blocked when mode Each major _ refueling but 3/3/71 selector switch is in shutdown-not less frequently than position. once every twelve months. -Other Liquid poison system component Two months or less. 8 71, ~ -check.~ 10 71 and 12 71 Post-incident spray system At each major refueling 3/3/7L

automatic control operation.-

shutdown but not less frequently than'once a year. ' '(

36 System or Function Frequency of Dates Routine Tests Tested Undergoing Test Other (Contd) Core spray system trip circuit. Not less frequently than 3/10/71 once every twelve months. Emergency condenser trip Not less frequently than 3/12/71 circuits. once every twelve months. Containment Containment sphere access air Once every six months or 10/11/71 locks and vent valves, leakage less. rate. Isolation valve operability and At least once overy 3/4/71 leak tests. twelve months. Isolation valve controls and Approximately quarterly. 10/11/71 instrumentation tests. Penetration inspection. At least once every 3/2/71 twelve months. Integrated leak test. Once every two years. 3/25/70 The following instrument checks and calibrations were performed L at least once a :aonth: 1. Reactor safety system checks not requiring plant shutdown. I 2. Air ejector off-gas monitor. 3 Stack-gas monitor calibration. 4. Emergency condenser vent monitor. l. -5 Process monitor. 6. Area monitoring system. CONSUMERS POWER C EPANY By: NI v Ralph B. Sewell / Nuclear Licensing Administrator Date: February 28, 1J72 f Sworn and subscribed to before me this 28th day of February 1972. I Y C O'LOQqnm c >t_I h 'AJ Notary Public, Jackson County,. Michigan My commission expires December 8, ISr/5 .}}

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