ML20214L314

Semiannual Radioactive Effluent Release Rept,Jan-June 1986

ML20214L314
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/30/1986
From:	Gahm J PUBLIC SERVICE CO. OF COLORADO
To:	Gagliardo J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
References
P-86534, NUDOCS 8609100077
Download: ML20214L314 (21)

Text

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SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT January - June 1986 Public Service Company of Colorado Fort St. Vrain Nuclear Generating Station August 1986 l

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l 8609100077 860630 PDR ADOCK 05000267 lEIV l

_ - . - _ _ _. _. Ih

1.0 Summary This report summarizes the radiological effluent released from the Fort St. Vrain Nuclear Generating Station for the period January through June, 1936. This information is provided pursuant to the requirements of Sections 7.5.1.e, 8.1.1.g.8), 8.1.2.d, and 8.2.1.h.1) of the Fort St. Vrain Technical Specifications.

An attempt has been made during this report period to follow the report format recommended by Regulatory Guide 1.21 as well as the requirements of the aforementioned sections of our technical specifications.

Along with a supplemental information sheet the following tables are included with this repo-t:

! Table Description IA Gaseous Effluents - Summation of All Releases 1C Gaseous Effluents - Grcund-Level Releases 2A Liquid Effluents - Summation of All Releases 28 Liquid Effluents 3 Solid Waste and Irradiated Fuel Shipments Please note that Table IB (of Regulatory Guide 1.21) has been omitted from this report due to the fact that all of our gaseous effluents are assumed to be ground-level releases as opposed to elevated releases. In addition, Table 4A (of Regulatory Guide 1.21) which lists meteorology data collected for periods of continuous and batch gaseous releases, has also been omitted due to problems associated with the computer summarization of these data. The meteorology data will be transmitted when it becomes available as an addendum to this report.

Fort St. Vrain Technical Specifications apply exclusively to the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions.

This list does not mean that only these nuclides are considered.

Other gamma emitting nuclides that are identifiable, together with those of the above nuclides, are analyzed and reported in this report.

~-

The lower limit of detection (LLD), for the purposes of this report, is defined as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a real signal.

The LLD values specified in our Technical Specifications are as follows:

Liquid Principle Gamma Emitters 5.00E-07 pCi/mi Dissolved Noble Gases 1.00E-05 pCi/mi Tritium 1.00E-05 pCi/ml Iodine-131 1.00E-06 pCi/ml Gross Alpha 1.0GE-07 pCi/ml Strontium-89, 90 (Composite) 5.00E-08 pC1/ml Gaseous Principle Gamma Emitters 1.00E-04 pCi/cc (Gas)

Principle Gamma Emitters 1.00E-11 pCi/cc (Particulate)

Tritium (Gas) 1.00E-06 pCi/cc Iodine-131 (Charcoal) 1.00E-12 pCi/cc Gross Alpha (Particulate) 1.00E-11 pCi/cc Strontium-89, 90 (Particulate) 1.00E-11 pCi/cc l

Gross-Beta (Particulate) 1.00E-11 pCf/:c Where applicable, we have listed "less-than" values for those nuclides listed specifically in our Technical Specifications. These "less-than" values were calculated using the observed LLD values and the total volume of the media. "Less-than" values were not included in the total values for pathway.

On Table 1A the percent of technical specification limit is blank because this value could not be calculated from data which were at or below the minimum detectable activity. On Table IC, the continuous release mode values cannot be determined because this release pathway is the same as the batch mode. All other blanks on Tables 1C and 28 are due to the fact that no LLD values for these nuclides are required to be calculated per technical specifications.

In the past there has been some confusion as to the total volume of water used for dilution of radioactive liquid effluent. All average diluted concentrations are based on the activity at the unrestricted area. Although this effluent could eventually reach one of two rivers (St. Vrain Creek and South Platte River) which converge approximately one and one half mile downstream of the plant no further dilutions were assumed. Additional discussi3 ,n river flow is contained in section 4a) of the Supplemental Informson Report.

During this report period there were no unplanned radioactive liquid waste releases.

On April 3, 1986, an unplanned gaseous release occurred as a result of a gaseous release into the Reactor Building. The gas was released through the normal Reactor Buildir.g exhaust ventilation system. This activity was detected and recorded by the noble gas monitor RT-7324-1 which increased from 180 to 250 counts per minute. Based on the exhaust flowrate, monitor sensitivity, release duration, and laboratory isotopic results, the total activity released was calculated to be 9.6E-02 curies of ::enon-135. Based on meteorology data the concentration at the exclusion area boundary was calculated to be 3.4E-11 pCi 135Xe/cc which is approximately 3000 times less than the MPC specified in 10CFR20. A complete description of the event can be found in Licensee Event Report 86-017.

There were no radioactive solid waste shipments made during this report period.

There were no major changes made to the radioactive waste systems during this report period.

There were no changes made to the Process Control Program (SUSMAP-3) during this report period.

There was one change made to the Offsite Dose Calculation Manual (SUSMAP-2). Issue 12, which became effective on April 18, 1986, changed the sample location of an adjacent milk sampling facility.

This was necessary because the previous dairy did not wish to continue in the voluntary sampling program. Another dairy in the same general area was then recruited for this program. No other changes were made during this report period.

i 1

During this report period one Reactor Building ventilation exhaust stack monitor was inoperable for a period of greater than 30 days.

This monitor, RT-73437, is a particulate and iodine monitor manufactured by Eberline Instrument Company. It was placed out of service on April 3, 1986, due to inadequate flow through the monitor.

This monitor remained out of service for the duration of the report period due to non-availability of repair parts. During the period of inoperability the gaseous effluent from the Reactor Building was continuously monitoted and recorded by monitors RT-4801 (particulate) and RT-4802 (iodine). The inoperability of this monitor was of no radiological significance due to the redundancy of our stack monitoring equipment.

Per our Technical Specifications the Radioactive Effluent Release Report submitted within 60 days after January 1, of each year will include an assessment of radiation doses to show conformance with 40CFR Part 190 for the entire year; therefore the radiation doses are not included in this report.

Regarding the total release value -l gross alpha listed in section D of Table 2A it should be pointed out that this value is obtained by multiplying the average value, of which most observations are at or below the minimum detectable activity (MDA), by an extremely larger volume (> 1.00E+07 liters). This may lead the reader to the erroneous conclusion that the liquid effluent contained significant alpha activi ty , when, in fact, there has been no measurable alpha activity observed anywhere in the plant. To address this problem we will perform gross alpha determinations on monthly composites using equipment that is more sensitive to alpha radiation and will thus have a much lower MDA.

. . TABLE 1A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1986 GASEOUS EFFLUENT-SUMMATION OF ALL RELEASES l Unit l Quarter l Quarter l Est. Total l l l 1 l 2 l Error. % l l l l l l

~A. Fissien and activation products

1. Total release l C1 l 4.31E+00 l 5.14E+01 l 1.70E+01 l l l l l 1

2. Average release rate for luci/secl 5.54E-01 [ 6.54E+00 l period l l l l l l l l

3. Percent of technical l  % l 5.29E-02 l 6.86E-01 l specification limit l l l l B. Iodine

1. Total iodine-131 l Ci l <3.87E-06 l <3.87E-06 1 0.00E+00 l 1 l l l l

2. Average release rate for luci/secl <4.98E-07 l <4.92E-07 l period l l l l l _l l l

3. Percent of technical l  % l l l specification limit l l l l C. Particulates

1. Particulates with half-lives l Ci l <1. '1E-07 l <1.71E-07 l 0.00E+00 l

> 8 days l l l l l 1 l l 1

2. Average release rate for luci/secl <2.20E-08 l <2.17E-08 l period l l l 1 l l l 1

3. Percent of technical l  % l l 1 specification limit l l l l l l l l

4. Gross alpha radioactivity l Ci l <5.69E-08 1 <5.09E-08 l 1 I I l D. Tritium

1. Total release l Ci l 2.70E+00 l 3.39E-01 l 1.70E+01 l l i l l l

2. Average release rate for luci/secl 3.47E-01 1 4.31E-02 l period l l l l l l l l

3. Percent of tech. spec. l  % l 2.63E-02 1 4.39E-03 l limit l I l l

TABLE IC EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1986)

GASEOUS EFFLUENTS--GROUND-LEVEL RELEASE CONTINUOUS MODE BATCH MCDE IReleased l Unit I Quarter 1 l Quarter 2 l Quarter 1 I Quarter 2 l l l l l l l l Fission gases Ikrypton-85 I Ci l l l l l l 1 I I I I I l krypton-85m I C1 I l l 1.20E-01 l 5.48E-01 l l l l l l 1 1 Ikrypton-87 l C1 I l l 3.64E-02 l 6.00E-01 I l- l l l l l l Ikrypton-88 l Ci l l l 1.18E-01 1 1.04E+00 l l 1 I I I I I Ixenon-133 I Ci I l l 3.30E+00 1 4.63E+01 l l 1 I I I I I lxenon-135 i Ci l 1 l 7.19E-01 1 2.17E+00 l I I I I I I I Ixenon-135m l Ci i l l I 6.71E-02 I I I I I I I I Ixenon-138 I Ci l l I <5.75E-01 I <3.98E-01 I I I I I I I I Ixenon-133m l Ci l I I 8.90E-03 I 3.31E-01 l l l l l l l l lxenon-131m I Ci l I I l 3.03E-01 I I I I I I I I l Total for period

• l Ci l l l 4.30E+00 l 5.14E+01 l l l l l l i I Iodines liodine-131 l Ci i I l l l l l l I I l l l iodine-133 I Ci l l l l l l 1 I I I I I liodine-135 l Ci I l 1 1 I I I I I I I I ITotal for period l C1 I l l l l l 1 I I I I I

• Total values do not include "<" data

Page 2 TABLEIC(Continued)

Particulates l strontium-89 l Ci l l l l l l l l l l l l l strontium-90 l Ci l l l l l l l l 1 1 I I l cesium-134 l Ci l l l l l l l l 1 1 I I l cesium-137 l Ci I l l l l l l l l l l 1 l barium-lanthanum-140 l Ci l l l l l l l l l l l l 4

f i

i I

Total valuat do not include "<" data

,_,,,,.,,,,__m...,_, _ _ _ , , _ _ . _ _ .

-- - - - . - - - , - - .,-_ m_ . , , -- ,e, , - , ...m___, ,,m , , , , . - ,g...,,%,, ..,y,_

. . TABLE 2A EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT 1986 LIQUID EFFLUENT-SUMMATION OF ALL RELEASES l Nuclide l Units l Quarter l Quarter l Est. Total l l l l 1 l 2 l Error. % l 1 l l l l l A. Fission and activation products

1. Total release lCs-137 l Ci l 1.86E-06 l 7.14E-06 l 1.70E+01 l l 1 I I I I

2. Average diluted luci/ml 1 3.44E-12 l 1.31E-11 l concentration l l l l l l l l

3. Percent of I  % l 3.05E-03 l 2.65E-03 l applicable limit l l l l

1. Total release lCo-60 l Ci l 2.20E-06 l 2.45E-06 l 1.70E+01 l 1 I I I I l-

2. Average diluted luCi/mi l 4.07E-12 l 4.49E-12 l concentration l l l l l l 1 l

3. Percent of l  % l 4.03E-03 l 2.31E-03 l applicable limit l l l l B. Tritium

1. Total release lH-3 l Ci l 1.98E+01 l 7.29E+01 l 1.70E+01 l l l l l l 1

2. Average diluted luc 1/ml l 3.67E-05 l 1.34E-04 l concentration l l l l l 1 I I

3. Percent of l  % l 2.05E-01 l 2.95E-01 l applicable limit l l l l C. Dissolved and entrained gases

1. Total release lXe-133 l Ci l 2.38E-04 l 1.66E-03 l 1.70E+01 l l l- 1 I I I

2. Averaga diluted juCi/ml l 4.41E-10 l 3.04E-39 l concentration l l l l 1 l l l

3. Percent of l  % l 8.65E-03 l 1.17E-02 l applicable limit l l l l D. Gross alpha radioactivity

1. Total release l Ci l 3.93E-04 l 3.33E-04 l 1.70E+01 1 l l 1 I I l

Page 2 TABLE 2A (Continued)

'r E. Volume of waste released (prior to dilution) l Liters l 1.86E+07 l 1.54E+07 l 1.00E+01 l 1 1 I I I l F. Volume of dilution water used during release ILiters l 5.40E+11 l 5.46E+11 l 1.00E+0) l i l I I I i 1

1

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1 a

~- . ., . - . - --. , . - - . - - - - - . - - - . . . - - ,- -. - _ -... . . - . - .

. . TABLE 28 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1986)

LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE INuclides Released l Unit l Quarter 1 l Quarter 2 l Quarter 1 l Quarter 2 l l l l 1 I I I l strontium-89 l Ci l <3.22E-05 1 <E.57E-05 l <4.86E-06 l <3.68E-06 I I I -l l I I I l strontium-90 1 Ci l <3.37E-05 l <2.69E-05 l <5.07E-06 l <3.86E-06 l l l l 1 1 I I l cesium-134 l Ci l <1.19E-03 l <9.75E-04 1 <1.61E-04 l <1.42E-04 l 1 I I I I I I l cesium-137 l Ci i <1.12E-03 l <9.13E-04 l 1.86E-06 l 7.14E-06 l l 1 I I I I I l iodine-131 l Ci l <9.49E-04 l <7.75E-04 l <1.28E-04 l <1.13E-04 l 1 I I I I I I l cobalt-58 l Ci l <1.13E-03 l <9.20E-04 l <1.52E-04 l <1.34E-04 l l 1 I I I I I l cobalt-60 l Ci l <1.10E-03 l <8.97E-04 1 2.20E-06 l 2.45E-06 l l l l 1 I I I l iron-59 l Ci l <1.94E-03 l <1.58E-03 l <2.61E-04 l <2.31E-04 l l l 1 I I I I l zinc-65 l Ci l <2.22E-03 l <1.81E-03 I <2.99E-04 l <2.64E-04 l l I I I I I I

, l manganese-54 l Ci l <1.03E-03 1 <8.41E-04 l <1.39E-04 1 <1.22E-04 l 1_ l l I I I I l chromium-51 l C1 l l l l l l l 1 I I I I l zirconium-niobium-95 l Ci l l l l l 1 1 I I I I I l molybdenum-99 l Ci l <7.07E-03 l <5.77E-03 l <9.52E-04 l <8.41E-04 I I I I I I I I l technetium-99m l C1 1 l l l l l l l l l 1 1 I barium-lanthanum-140 l Ci l l l l 1 I I I I I I I l cerium-141 l Ci l <1.54E-03 l <1.26E-03 l <2.07E-04 l <1.83E-04 l l 1 I I I I I l tritium l Ci l 3.51E-02 l 2.55E-01 l 1.98E+01 1 7.26E+01 l 1 I I I I I I I sulfur-35 l C1 l l l 3.05E-05 l 2.50E-04 1 I I I I I I I

! l Total for period (above)*l Ci l 3.51E-02 l 2.55E-01 l 1.98E+01 l 7.26E+01 1 I I I I I I I Total values do not include "<" data

Page 2 TABLE 28'(Continued) l xenon-133 ( Ci i <2.93E-03 l <2.39E-03 l .2.38E-04 l 1.66E-03 l l 1 -l I I i 1 l xenon-135 l Ci l <8.14E-04 l <6.65E-04 l <1.10E-04 l <9.68E-05 l l l l l 1 I i 1

i l

4 i

J d

1 i

l i

• Total values do not include "<" data

, - - - . , . . , ..-_..----,--m , , ,~,, , ,,.-mm.,.r,. . _ , , 3,.v.... ._,_m

. . TABLE 3 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1986)

SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (Not irradiated fuel) l 1. Type of waste l Unit I 6-month l Est. Total l l l l Period I Error. % l l l l l l l l l l .I l a. Spent resins, filter l m' l 0.00E+00 l l l . sludges, evaporator l Ci l 0.00E+00 l l l bottoms, etc. l l l l l l l l l l l l l 1 l b. Dry compressible waste, l m' l 0.00E+00 l l l contaminated equip. etc. l Ci l 0.00E+00 l l 1 I I I I I I I I I l c. Irradiated components, l m' l 0.00E+00 l l l control rods, etc. I Ci l 0.00E+00 l l l l l 1 1 I I I I I l d. Otaer(describe) l m' l 0.00E+00 l l l l Ci l 0.00E+00 l l l l 1 1 I

2. Estimate of major nuclide composition (by type of waste)

a. Not applicable I  % l l l  % l l l  % l l

b. I  % l l l  % l I I  % I i

c. I  % l l l  % l l I  % I I

d. I  % I I I  % l i I  % l l

3. Solid Waste Disposition Number of Shipments Mode of Transportation Des _tination 0

B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shipments Mode of Transportation Destination 0

Effluent and Waste Disposal Semiannual Report Supplemental Information Facility Fort St. Vrain Nuclear Generating Station Licensee Public Service Company of Colorado

1. Regulatory Limits All results of radioactivity analyses of gaseous and liquid effluent are used in accordance with the methodology and parameters listed in the Offsite Dose Calculation Manual (SUSMAp-2) to assure that the concentrations at the point of release are maintained within the limits set forth in the Technical Specifications. These limits will ensure the quantity of radioactive effluent released from the plant is maintained as low as reasonably achievable and in any event within the limits of 10CFR20 and in accordance with 10CFR50.

The air dose due to noble gases released in gaseous effluent at the unrestricted area is limited to:

a) 5 millirads gamma and 10 millirads beta during any calendar quarter, and,

' b) 10 millirads gamma and 20 millirads beta during any calendar year.

The dose to a member of the public due to I-131, tritium, and radioactive particulates with half-lives longer than eight days in gaseous effluents will be limited to:

a) 7.5 millirems to any organ during any calendar quarter, and, b) 15 millirems to any organ during any calendar year.

The dose rate due to radioactive gaseous effluent is limited to the following:

a) For noble gases, less than or equal te 500 millirems per year to the total body and less than or equal to 3000 millirems per year to the skin, and, b) For I-131, tritium, and radioactive particulates with half-lives greater than eight days, less than or equal to 1500 millirems per year to any organ.

. c The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released to unrestricted areas are limited as follows:

a) During any calendar quarter to le_ss than or equal to 1.5 millirem to the total body and to less than or equal to 5 millirems to any organ, and, b) During any calendar year to less than or equal to 3 millirems to the total body and to less than or equal to 10 millirems to any organ.

2. Maximum Permissible Concentrations All MPCs used in determining allowable release rates from the gas waste holdup system and the liquid waste system are those listed in Table II, Columns 1 and 2, respectively, of Appendix B to 10CFR20. In addition, for the MPC of dissolved noble gases in liquid effluent the value of 2.00E-04 microcuries per milliliter was used.

For the purposes of calculating allowable release rates, the MPC for halogens and particulates with half-lives longer than eight days will be reduced by a factor of 700 from their listed value in Table II, Column 1, of Appendix B to 10CFR20.

3. Average Energy The average energy (E-Bar) of the radionuclide mixture in releases of fission and activation gases is not calculated or used at this facility.

4. Measurements and approximations of Total Radioactivity a) Fission and Activation Gases Batch releases from the gas waste holdup system are performed after sampling and analyses for noble gases and tritium. These analytical results are used along with atmospheric dilution factors to determine the allowable release rate. Gas is released on a continuous basis through a gas waste header which is monitored by a noble gas monitor and an iodine monitor. In the event of high activity in the continuous release header control functions are initiated which divert the gas to the gas waste holdup system.

All radioactive gases are released to the Reactor Building exhaust ventilation system which has a flow rate of approximately 32,000 cubic feet per minute. The full-flow of this exhaust is directed through high efficiency particulate filters (HEPA) and activated charcoal beds prior to the release to the environment.

Downstream of the activated charcoal beds the gas stream radioactivity is continuously monitored and recorded by noble gas monitors, particulate monitors, and iodine monitors.

b) Iodines For gaseous iodine, the Reactor Building exhaust ventilation is monitored and recorded on a continuous basis. The 2-inch iodine cartridges used in these monitors are removed from service after one week of service and quantitatively analyzed on a gamma spectroscopy system. The quantity of radiciodine released during that period would be calculated based on the integrated flow during the collection period.

c) Particulates As in the case of iodine discussed in b) above, a 2-inch particulate filter is removed and analyzed each week.

Gross-beta analysis as well as gamma spectral analyses are performed to identify and quantify any radionuclides. The quantity of any radionuclides on this filter with half-lives greater than eight days would similarly be correlated to total flow during the collection period.

d) Liquid Effluents All liquid effluent discharged from the site reaches the unrestricted area at the Goosequil Ditch. From that point the effluent can be diverted to the St. Vrain Creek via the St. Vrain Slough, or, more commonly diverted to the Goosequil Pond which is approximately one mile north of the plant site. Outfall from the Goosequil Pond reaches the South Platte River. Both rivers converge approximatley 1 1/2 miles from the plant site. The average stream flow reported in section Sa) of this supplemental report is a summation of both rivers and was received and tabulated from data provided by the Colorado Department of Natural Resources in Greeley, Colorado.

Liquid effluent is released from the site using both a continuous and batch mode. The continuous mode (automatic discharge mode) is used on the Turbine Building Sump effluent where the only expected radionuclide is tritium.

This discharge path utilizes a continuous sampler and an aliquot of this composite sampler is taken three times per week and analyzed for gross-beta, gross-alpha, tritium, and gamma emitters. Total flow integrators enable us to calculate the total activity released via this pathway based on composite sample results. Discharge from the Turbine Building Sump is made directly to the unrestricted area with no dilution.

The batch release mode is used on the Reactor Building Sump effluent and the liquid waste processing system. The Reactor Building Stap area can hold several hundred thousand gallons of waste water from various sources which could be contaminated. Tte liquid waste system consists of 2-2000 gallon receivers, 1-2000 gallon monitoring tank, and associated filters and demineralizers. This system is designed to collect and process contaminated waste water resulting frori. reactor and laboratory operations.

Prior to etch liquid batch release duplicate samples are quantitatively analyzed for their radioactive constituents.

These analyses include gross-beta, gross-alpha, tritium, and gamma spectral analyses. The results of these analyses, and other analyses as dictated by the gross-beta results, are used to determine the maximum release rate from the site.

The liquid effluent, normally released at or less than 60 gallons per minute, is diluted with cooling tower blowdown, which runs at or more than 1100 gallons per minute. The resulting mixture is sampled during the release period to confirm compliance with regulatory limits.

The liquid effluent from the batch release mode is monitored continuously by redundant gamma activity monitors.

All tank level indicating devices, flow monitoring and recording devices, and radiation monitoring equipment are calibrated and maintained at scheduled intervals in accordance with established procedures.

Composite samples from batch releases and continuous releases are analyzed monthly for Sr-89 and Sr-90.

Composite samples from batch releases from the liquid waste

, processing system are analyzed monthly for S-35. All sample results are conservatively decay-corrected to the start of the composite period.

e >

l e) Overall Errors The overall error associated with determining the total activity released from the site for both gaseous and liquid effluent is estimated to be 17.3 percent. This value is the square root of the sum of squares of counting statistics and associated calibration errors, sampling errors, and tank volume estimates, each considered to be plus or minus 10 percent.

5. Batch Releases a) Liquid i I l l Number of Batch Releases l 8.90E+01 l l Total Time Period for l l 1 Batch Releases l 6.04E+02 HOURS l l Maximum Time Period for l l la Batch Release 1 4.06E+01 HOURS l l Average Time Period for l l la Batch Release 1 6.78E+00 HOURS l l Minimum Time Period for l l la Batch Release i 1.27E+00 HOURS l l Average Stream Flow During l l l Periods of Release of l l l Effluent into a Flowing l l IStream l 3.01E+05 GPM* l

• Gallons Per Minute b) Gaseous I l l INumber of Batch Releases l 1.92E+02 l l Total Time Period for l l IBatch Release 1 8.34E+02 HOURS I l Maximum Time Period for l l l Batch Release l 1.03E+01 HOURS I l 1 l l Average l 4.34E+00 HOURS l l l l l Minimum i 1.00E+00 HOURS I

r.

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• s' I

6. Abnormal Releases a) Liquid I I I l Number of Releases l 0.00E+00 l l l l l Total Activity Released 1 0.00E+00 CURIES l b) Gaseous l l l l Number of Releases i 1.00E+00 l l l 1 ITotal Activity Released l 9.60E-02 CURIES I

l O PublicService . ._i. .

Company of Colorado 16805 WCR 19 1/2, Platteville, Colorado 80651-9298 August 25, 1986 Fort St. Vrain Unit #1 P-86534 Regional Admini:+rator, Region IV 3ggg~g]%

U. S. Nuclear Regulatory Commission -

611 Ryan Plaza Drive, Suite 1000 is 1

Arlington, Texas 76011 SEP - 2 E Attn: Mr. J. E. Gagliardu, Chief ;U Reactor Projects Branch g Docket No. 50-267

SUBJECT:

Semi-Annual Radioactive Effluent Release Report

Dear Mr. Gagliardo:

Attached please find the Semi-Annual i<adioactive Effluent Release Report for the Fort St. Vrain Nuclear denerating Station.

This report covers the period January 1,1986 through June 30, 1986, and is submitted pursuant to Section 7.5.1 e. of the Fort St. Vrain Technical Specifications.

Please contact Mr. Mike Holmes at (303) 480-6960 if you have eny questions regarding this report.

Sincerely, jhTYh J. W. Gahm Manager, Nuclear Production Fort St. Vrain Nuclear Gmerating Station JWG:VJM/skd Attachments

%-ssi y, YQ

9 8 cc w/ attachments:

Dottie Sherman, ANI Library American Nuclear Insurers The Exchange Suite, 270 Farmington Avenue Farmington, Connecticut 06032 Mr. Albert J. Hazle, Director Radiation Control Division Colorado 'epartment of Health 4210 E. lith Ave.

Denver, CO 80220 Mr. Mike Dolphin, Site Manager General Atomic Company P.O. Box 426 Platteville, CO 80651 Mr. Paul B. Smith Regional Representative, Radiation Program U.S. Environmental Protection Agency Region VIII Suite 900, 1860 Lincoln St.

Denver, CO 80203 Mr. Frank Rozich, Director Water Pollution Control Division Colorado Department of Health 4210 East lith Ave.

Denver, CO 80220 Mr. George Campbell Environmental Science Division

, Rockwell International 123 Building P. O. Box 464

. Golden, CO 80401 1

Dr. James E. Johnson Radiology & Radiation Biology Dept.

. Colorado State University l Fort Collins, CO 80523 i

Mr. Robert l'arrell

' Senior NRC Resident Inspector Fort St. Vrain Station