ML20212E156
| ML20212E156 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/31/1986 |
| From: | Robert Williams PUBLIC SERVICE CO. OF COLORADO |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| P-87078, NUDOCS 8703040281 | |
| Download: ML20212E156 (46) | |
Text
{{#Wiki_filter:- ____ _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ . SEMI-ANNUAL RADI0 ACTIVE EFFLUENT-RELEASE REPORT July - December 1986 Public Service Company of Colorado Fort St. Vrain Nuclear Generating Station February 1987 8703040281 861231 PDR ADOCK 05000267 R PDR I I
1.0 Summary This report summarizes the radiological effluent released from the Fort St. Vrain Nuclear Generating Station for the period July through December, 1986. 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 and 40CFR190, subpart B. The following tables with a supplemental information section are included with this report: Table Description IA Gaseous Effluents - Summation of All Releases 1C Gaseous Effluents - Ground-Level Releases 2A Liquid Effluents - Summation of All Releases 2B Liquid Effluents 3 Solid Waste and_ Irradiated Fuel Shipments 4A Hourly Meteorological Data 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. 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 included 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/ml Tritium 1.00E-05 pC1/ml Iodine-131 1.00E-06 pCi/ml Gross Alpha 1.00E-07 pCi/ml Strontium-89,90(Composite) 5.00E-08 pCi/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 Gross-Beta (Particulate) 1.00E-11 pCi/cc 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.
The percent of technical specification limit on Table 1A 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 2B are due to the fact that no LLD values for these nuclides are required to be' calculated per technical specifications. -There has been some confusion as to the total volume of water used-for dilution of a 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 discussion on river flow is contained in the Supplemental Information Section. No unplanned radioactive liquid waste or gas waste releases were made during this report period. There were no radioactive solid waste shipments made during this report period. There were no changes to the Process Control Program (SUSMAP-3) during this report period. Two changes were made to the Offsite Dose Calculation Manual, SUSMAP-2. Issue 13, effective date September 30, 1986, changed the set point units designator from cpm /pC1/cc to cpm /pCi. Issue 14,
-effective date January 21, 1987, changed the set point calculations for RT-7325-1, RT-73437-1, RT-4802 and RT-6314-1 to include a 0.9 factor to account for the 90 percent retention efficiency of iodine on a silver ' zeolite cartridge. Also, a note was moved to the beginning of a step and all references to the word " instrument" were changed to " monitor."
Meteorological data collected for the period July through December, 1986 is included as an addendum to this report in the format of Table 4A of Regulatory Guide 1.21. Meteorological data collected for the period January through June 1986 was submitted to the Nuclear Regulatory Commission via Public Service Company letter P-87028.
Radiation doses resulting from the release of radioactive liquid and gaseous effluents from Fort St. Vrain during 1986 are reported below. Radiation doses are calculated in accordance with the Fort St. Vrain Offsite Dose Calculation Manual (Procedure SUSMAP-2), which is based on NUREG-0472, " Radiological Effluent Technical Specifications for PWR's," NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," and other inputs from the Nuclear Regulatory Commission (NRC). The NRC recently reviewed the Fort St. Vrain Offsite Dose Calculation Manual (0DCM) with the assistance of the Idaho National Engineering Laboratory and determined that "
... the Fort St. Vrain ODCM generally uses documented and approved methods consistent with the guidelines in NUREG-0133." As it is not practical to list all of the assumptions used in making the dose assessment to the public, the reader is directed to SUSMAP-2 for further details.
Doses are calculated for a hypothetical " maximum" individual present at all times of the year in the sector with the highest X/Q and D/Q values. Liquid - 10CFR50 Whole Body 1.4E-1 mrem Organ (maximum) 1.5E-1 mrem Gas - 10CFR50 Noble Gas Gamma 5.3E-3 mrad Beta 8.9E-3 mrad Iodine, Particulates, Tritium Adult Whole Body 2.3E-3 mrem Organ (maximum) 2.3E-3 mrem Teen Whole Body 2.9E-3 mrem Organ (maximum) 2.9E-3 mrem Child Whole Body 4.3E-3 mrem Organ (maximum) 4.3E-3 mrem Infant Whole Body 3.2E-3 mrem Organ (maximum) 3.2E-3 mrem
Gas - 10CFR20 Iodine, Particulates, Tritium 2.7 mrem The total dose for 1986 was less than 3 mrem including contributions from liquid and gaseous releases. It should be noted that the dose due to gaseous releases was calculated using a X/Q of 3.65E-6 s/m'. This is the value obtained from meteorological data collected at Fort St. Vrain for the period December 1984 through November 1985. It is felt that use of this dilution factor is more accurate than the dnnual average X/Q of 1.37E-6 s/m' as previously reported in the FSAR. Beginning with 1987, dilution factors for actual periods of release will be used to calculate doses from gaseous effluent releases. The doses from gaseous releases are calculated including contribution from direct radiation; a review'of TLD data collected in 1986 has confirmed that no measurable direct radiation exposure is attributable to Fort St. Vrain. As mentioned earlier, the doses reported here are to the hypothetical most exposed member of the public. In order to assess the actual dose to the likely most exposed member of the public due to their activities inside the site boundary, the following assumptions are made:
- Residents living within the site boundary are at home during 50% of gaseous effluent releases.
- As game fishing is not prevalent within the site boundary, fish consumption is 25% of the adult fish consumption of 21 kg/yr as listed in SUSMAP-2.
- All other assumptions of SUSMAP-2 remain valid.
The resultant maximum doses are as follows: Gas 1.4 mrem Liquid 3.5E-2 mrem The doses demonstrate conformance with the exposure limit in 40CFR Part 190 of 25 mrem to the total body, 75 mrem to the thyroid, and 25 mrem to any organ.
To show conformance with 40CFR190 subpart B, the total curies of Krypton-85 released is <2.61E+01. The 29.78 key iodine-129 peak is below the minimum detectable energy of our detectors. We assume that no iodinc-129 (fission yield of 0.574 percent) has been released because no measurable amount of iodine-131 (fission yield of 2.78 percent) has been detected at this station. The total release value of gross alpha is listed in Table 2A Section D. We are currently in a six month program to evaluate an online liquid beta monitor installed in the plant via Change Notice 2313. Details of the installation were submitted to the Nuclear Regulatory Commission via Public Service Company letter P-86655. During the report period, six (6) Reactor Bufiding ventilation exhaust stack monitors were inoperable for a period of greater than 30 days. RT-4801 is a particulate monitor and was out of service for 45 days awaiting test equipment needed for calibration. RT-4802 is an iodine monitor which was also out of service for 45 days awaiting test equipment needed for calibration. RT-4803 is a noble gas monitor which was out of service for the entire reporting period awaiting test equipment needed for calibration. These monitors were discussed in Licensee Event Report (LER) 86-023. RT-7325-1 is an iodine monitor and was out of service for 158 days awaiting test equipment needed for calibration. RT-73437-1 and 73437-2 are iodine and particulate monitors respectively and were out of service for 68 days due to equipment problems, lack of repair parts, and lack of test equipment needed for calibration. An annual land use census is performed as part of the Fort St. Vrain Radiological Environmental Monitoring Program. As it is currently our practice to calculate radiation doses using the annual average X/Q and D/Q values for Fort St. Vrain, changes in the land use census do not have any impact on calculated doses. Changes made to environmental sampling locations as a result of the annual land use census are reported in the annual Radiological Environmental Monitoring Program (REMP) report.
SUPPLEMENTAL INFORMATION Facility: Fort St. Vrain Nuclear Generating Station Licensee: Public Service Company of Colorado
- 1. Regulatory Limits 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 that the quantity of radioactive effluent released from the plant is maintained as low as reasonably achievable and 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 4 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 effluent is 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:
- a. For noble gases, less than or equal to 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.
The dose or dose commitment to a member of the public from radioactive material in liquid effluent released to unrestricted areas are limited as follows:
- a. During any calendar quarter to less than or equal to 1.5 millirem to the total body and to less than or equal to 5 millirems to any organ, and, t
(
~2-
- 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 ary organ.
- 2. Maximum Permissible Concentrations All MPC values in determining allowable release rates from the gas waste system and the liquid waste system are those listed in Table II, Columns 1 and 2 respectively, of Appendix B to 10CFR20.
For the MPC of dissolved noble gases in liquid effluent, the value of 2.00E-04 microcuries per milliliter was used.
- 3. Average Energy The average (E-BAR) of the radionuclide mixture in release of fission and activation gases is not calculated nor used at this facility.
- 4. Measurements and Approximation of Total Radioactivity
- a. Fission and Activation Gases Batch releases from the gas waste 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 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 is directed through high efficiency particulate i 1 filters (HEPA) and activated charcoal beds prior to the i release to the environment.
l Downstream of the activated charcoal beds, the gas stream radioactivity is continuously monitored and recorded by noble gas monitors, particulate monitors, and iodine monitors. t %
6
- b. Iodine For gaseous iodine, the Reactor Building exhaust ventilation is monitored and recorded on a continuous bases. The iodine cartridges used in these monitors are removed from service after one week of service and quantitatively analyzed on a j gamma spectroscopy system. The quantity of radiodine
~,
- released during that period is calculated based on the integrated flow during the collection period.
. c. Particulate i As in the case of the fodine discussed in b. above, a f particulate filter is removed and analyzed each week. Gross alpha and gross beta analysis as well as gamma spectral analyses are performed to identify and quantify any radionuclides. The quantity of any radionuclide on this filter with half-lives greater than eight days would similarly be correlated to total flow during the collection period.
- d. Liquid Effluent All liquid effluent discharged from the site reaches the unrestricted area at the Goosequill 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 Goosequill Pond which is approximately one mile North of the p' mt site. Outfall from the Goosequill Pond reaches the Souv- Platte River. Both rivers converge approximately one and one half 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 calculated 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 for 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.
r-The batch release mode is used on the Reactor Building Sump effluent and the liquid waste processing system. The Reactor Building Sump area can hold several hundred thousand gallons of waste water from various sources which could be contaminated. The liquid waste system consists of two 2000 gallon receivers, one 2000 gallon monitoring tank and associated filters and demineralizers. This system is designed to collect and process contaminated waste water resulting from reactor operations. Prior to each release, duplicate samples are quantitatively analyzed for their radioactive constituents. These analyzes include gross beta, gross alpha, tritium and gamma spectral analyses. The result 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 flows at a minimum of 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 monitor equipment are calibrated and maintained at scheduled intervals in accordance with established procedures. Composite samples from the batch releases, and continuous releases are analyzed monthly for Sr-89, Sr-90 and S-35. All sample results are conservatively decay-corrected to the start of the composite period.
- e. Overall Error 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, associated calibration errors, sampling errors and tank volume estimates, each considered to be plus or minus 10 percent.
- 5. Batch Releases
- a. Liquid l Number of Batch Releases l 94 l l l l l Total Time Period For l l l Batch Releases l 9.52E+02 HOURS l l l l l Maximum Time Period For l l t la Batch Release. l 6.59E+01 HOURS l l l l l Average Time Period For l l la Batch Release l- 1.01E+01 HOURS l l l l l Minimum Time Period For l l i
la Batch Release l 2.50E-01 HOURS ~ l l l l l Average Stream Flow During l l l Period of Release of l l l Effluent Into a Flowing l l l Stream l 3.12E+05 GPM* l l 1 1
- Gallons Per Minute
- b. Gaseous l Number of Batch Releases l 23 l l l l
- l Total Time Period For l l.
l l Batch Releases l 1.01E+02 HOURS l l 1 I l Maximum Time Period For l l la Batch Release l 7.53E+00 HOURS l l l l l l Average Time Period For l l la Batch Release. l 4.39E+00 HOURS l l l l l Minimum Time Period For l l la Batch Release l 1.92E+00 HOURS l l l l r w-.- . . - - - - - - - - ,+ . - - . . . - __m - - _ _ , - - . . ._,.-
- 6. Abnormal Releases
- a. Liquid l Number of Releases l 0.00E+00 l l l l l Total Activity Released l 0.00E+00 CURIES l l l l
- b. Gaseous l Number of Releases l 0.00E+00 l l l l l Total Activity. Released l 0.00E+00 CURIES. l 1 1 I e
i . - - - - - . _ . . _
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 3 l 4 I Error. % l l l l l l A. Fission and activation products
- 1. Total release l Ci l 0.00E+00 l 0.00E-00 l l l l l l l
- 2. Average release rate for luc 1/secl 0.00E+00 l 0.00E-00 l Period l l l l l l l l
- 3. Percent of technical l % l 0.00E+00 l 0.00E-00 l specification limit l l__ l l B. Iodine
- 1. Total iodine-131 l Ci l <5.25E-06 l <5.25E-06 l l l l l l l
- 2. Average release rate for luc 1/secl <6.60E-07 l <6.60E-07 l period l l l l l l l l
- 3. Percent of technical l % l l l specification limit l l 1 l C. Particulates
- 1. Particulates with half-lives l Ci l <2.66E-07 l <2.51E-07 l l
> 8 days l l l l l l l 1 l
- 2. Average release rate for luCi/secl <3.34E-08 l <3.16E-08 l Period l l l l l 1 I i
- 3. Percent of technical l % l l l specification limit l l l l l l l 1
- 4. Gross alpha-radioactivity l Ci l <5.17E-08 l <4.91E-08 l l 1 I I D. Tritium
- 1. Total release l Ci l 5.46E-01 l 5.31E-03 l 1.70E+01 l l l l l l
- 2. Average release rate for luci/secl 6.87E-02 l 6.68E-04 l period l l l l l l l l
- 3. Percent of tech, spec. l % l 1.14E-03 l 1.11E-05 l limit l l l l
TABLE IC EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1986) GASEOUS EFFLUENTS--CROUND-LEVEL RELEASE CONTINU0US MODE BATCH MODE IReleased I Unit l Quarter I Quarter i Quarter 3 I Quarter 4 I I I I I I I I Fission gases ikrypton-85 I Ci l l l 0.00E+00 l 00.0E+00 I I I I I I I I ikrypton-85m I Ci l l I 0.00E+00 1 00.0E+00 l l l 1 l l l l Ikrypton-87 I Ci I l l <1.19E-02 l <6.97E-03 l l l l l l l l Ikrypton-88 I C1 l l l <1.83E-02 I <1.07E-02 I I I I I I I I -Ixenon-133 i Ci I I l <1.78E-02 1 <1.04E-02 l l 1 I I I I I lxenon-135 l Ci I l l <5.13E-03 l <3.01E-03 l l l l l l 1 I lxenon-135m l Ci l l l 0.00E+00 1 0.00E+00 I I I I I I I i lxenon-138 I C1 I I l <6.90E-02 l <4.04E-02 I I I I I I I I Ixenon-133m I Ci l l 1 <4.14E-02 l <2.43E-02 l l I I I I I I lxenon-131m l Ci l l l 0.00E+00 l 0.00E-00 I I l- 1 I I I 1 l Total for period
- I Ci l l l 0.00E+00 1 0.00E+00 l I I I I I I I Iodines liodine-131 l Ci I i l <3.15E-10 l <1.85E-10 I l - -- 1 I I I I I liodine-133 I Ci l l l 0.00E+00 1 0.00E+00 l 1 I I I I I i l1odine-135 I Ci l l l 0.00E+00 l 0.00E+00 l l l 1 I I I I ITotal for period l C1 l l l 0.00E+00 1 0.00E+00 l l l l 1 1 I I Total values do not include "<" data
Page 2 TABLE IC (Continued) Particulates l strontium-89 I Ci l l l l l e l l 1 l i I I l strontium-90 l Ci l l l l l l l l l l l l Icesium-134 I Ci I l l l l l 1 l l l l l ~ l cesium-137 l- Ci l l l l l l i I I I I i l barium-lanthanum-140 I Ci I I l l l 1 l l l l l l
- Total values do not include "<" data
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 3 l 4 l Error. % l l l l l l l A. Fission and activation products
- 1. Total release lCs-137 l Ci l 4.48E-06 l 1.98E-06 l 1.70E+01 l l l l l l l
- 2. Average diluted luc 1/ml l 8.12E-12 l 3.59E-12 l concentration l l l l l l l l
- 3. Percent of l % l 4.06E-05 l 1.80E-05 l applicable limit l l l l
- 1. Total release 100-60 l C1 l 2.49E-06 l 3.77E-07 l 1.70E+01 I I I l l I I
- 2. Average diluted luci/ml l 4.51E-12 l 6.83E-13 l concentration l l l l l l l l
- 3. Percent of l % l 1.50E-05 l 2.28E-06 l applicable limit l l l l B. Tritium
- 1. Total release lH-3 l Ci l 3.08E+01 l 3.80E+00 l 1.70E+01 l l l 1 I I I
- 2. Average diluted luci/ml l 5.58E-05 l 6.89E-06 l concentration l l l l l l l 1
- 3. Percent of l % l 1.86E+00 l 2.30E-01 l applicable limit j l l l C. Dissolved and entrained gases
- 1. Total release lXe-133 l Ci l 4.09E-06 l 0.00E+00 l 1.70E+01 l l l l l l l
- 2. Average diluted luCi/ml l 7.42E-12 l 0.00E+00 l concentration l l l l l 1 I l
- 3. Percent of l % l 3.71E-06 l 0.00E+00 l applicable limit l l l l D. Gross alpha radioactivity
- 1. Total release l Ci l<9.23E-05 l <3.98E-05 l 1.70E+01 l l l l 1 I l -_ _.
TABLE 2A (Continued) E. Volume of waste released (prior to diletion) l Liters l 1.57E+07 l 1.33E+07 l 1.00E+01 l 1 1 I I I F. Volume of dilution water used during release l Liters l 5.52E+08 l 5.52E+08 l 1.00E+01 l l l l l l f I
TABLE 2B EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1986) LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE lNuclides Released l Unit l Quarter 3 l Quarter 4 l Quarter 3 l Quarter 4 l l l l l l l l l strontium-89 l C1 l <3.83E-05 l <2.09E-05 l <9.54E-06 l <6.94E-06 l l l l l l l l l strontium-90 l Ci l <3.94E-05 l <2.19E-05 l <9.84E-06 l <6.74E-06 l l l l l l l l l cesium-134 .l Ci l <8.59E-04 l <7.22E-04 l <2.27E-04 l <2.04E-04 l l l 1 I I I I
.l cesium-137 l Ci l <8.32E-04 l <6.98E-04 l <2.09E-04 l <1.93E-04 l l l l 1 I I I l . iodine-131 l C1 l <7.12E-04 l <5.98E-04 l <1.88E-04 l <1.69E-04 l I I I I I I I l cobalt-58 l Ci l <7.49E-04 l <6.30E-04 l <1.99E-04 l <1.77E-04 I I I _I I I I I l cobalt-60 l Ci l <9.72E-04 l <8.16E-04 l <2.41E-04 l <2.23E-04 l l l I I I I I l iron-59 l Ci i <1.50E-03 l <1.26E-03 l <3.97E-04 l <3.55E-04 l l l 1 I I I I l zinc-65 l Ci l <1.73E-03 l <1.46E-03 l <4.59E-04 I <4.11E-04 l l l l l l 1 l l manganese-54 l Ci l <7.85E-04 l <6.60E-04 l <2.08E-04 l <1.86E-04 l l l l l l l l l chromium-51 l Ci l l l l l l 1 I I I I I l zirconium-niobium-95 l Ci l l l l l l l l l l 1 I l molybdenum-99 l Ci l <5.48E-03 l <4.60E-03 l <1.45E-03 l <1.30E-03 l l- 1 I I l l 1 l technetium-99m l C1 l l l l l I I I I I I I l barium-lanthanum-140 l C1 l l l l l l l 1 I I I I l cerium-141 l Ci l <1.17E-03 l <9.86E-04 [ <3.11E-04 l <2.78E-04 l l l 1 I I I I l . tritium l Ci l 8.08E-01 l 2.07E-01 l 3.00E+01 l 3.59E-00 l l 1 I I I I I l sulfur-35 l Ci l <2.55E-03 l <2.00E-03 l <3.29E-03 l <5.56E-04 l l l l l l l l l Total for period (above)*] Ci l 8.08E-01 l 2.07E-01 l 3.00E+01 l 3.59E-00 l l l l l 1 1 I
- Total values do not include "<" data
r L l Page 2 TABLE 28 (Continued) Continuous Mode Batch Mode l xenon-133 l Ci l <2.32E-03 l <1.95E-03 l <5.74E-04 l <5.48E-04 l l l l l l l~ l
-l xenon-135 l C1 - l <6.54E-04 l <5.50E-04 l <1.73E-04 l <1.54E-04 l l l l l l l l
- Total values do not include "<" data
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 l 6-month l Est. Total l l l l Period l Error. % l l l l l l l l 1 1 I l a. Spent resins, filter l m2 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 I I I I I l 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 l l 1 I I I I I I I l c. Irradiated components, l m8 l 0.00E+00 l l l control. rods, etc. l Ci l 0.00E+00 l l l l l 1 1 1 I I I I l d. Other(describe) l m' l 0.00E+00 l l l l Ci l 0.00E+00 l l l l l l l
- 2. Estimate of major nuclide composition (by type of waste)
- a. Not applicable I % l l l % l -1 i % I l
- b. I % l 1 l % l I l % l l
- c. I % l l i % l l l % l l
- d. I % I I I % 1 I l % i I
- 3. Solid Waste Disposition Number of Shipments Mode of Transportation Destination 0
B. IRRADIATED FUEL SHIPMENTS (Disposition) Number of Shipments Mode of Transportation Destination 0 l
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SUBJECT:
Semi-Annual Radioactive Effluent Release Report Gentlemen: Attached please find the Semi-Annual Radioactive Effluent Release Report for the Fort St. Vrain Nuclear Generating Station. This report covers the period July 1,1986 through December 31, 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 any questions regarding this report. Sincerely, Rowdliewsp~r by pgr R. O. Williams, Jr. Vice President, Nuclear Operations Fort St. Vrain Nuclear Generating Station R0W:FJB/djc Attachments S$ Ifl g
m 3 P-87078 February 28, 1987 cc w/ attachments: Regional Administrator, Region IV U. S. Nuclear Regulatory Commission Attn: Mr. J. E. Gagliardo, Chief Reactor Projects Branch 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Dottie Sherraan, ANI Library American Nuclear Insurers + The Exchange Suite, 270 Farmington Avenue Farmington, Connecticut 06032 Mr. Albert J. Hazle, Director Radiation Control Division Colorado Department 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 R,ckwell International
' 3 Building
- r. O. Box 464 Golden, CO 80401
, - - - , - - - , , ,.g - - - -- , -- ,, -7n-, ,- g- v.-,--, _,. ,,, - gw.-,--n- -
P-87078 February' 28, 1987 Dr. James E. Johnson Radiology & Radiation Biology Dept. Colorado State University Fort Collins, CO 80523 Mr. Robert Farrell Senior NRC Resident Inspector Fort St. Vrain Station i. l l l l'- I
- . . - - - , _ ,. _ _ _ . _ _ _ _ _ . . _ _ _ _ . _ _ _ _ _ , _ _ . . _ . _ _ _ _ _ . . _ _ _ _ . _ - -__ _ _ ___._.__}}