Text

Semiannual Radioactive Effluent Release Rept for Jan-June 1987
	ML20237J056
Person / Time
Site:	Fort Saint Vrain
Issue date:	06/30/1987
From:	Robert Williams; PUBLIC SERVICE CO. OF COLORADO
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	P-87297, NUDOCS 8709030476
	Download: ML20237J056 (32)
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SEMI-ANNUAL'RADI0 ACTIVE EFFLUENT '; '

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1.0 Summarv l This report summarizes the radiological effluent released from the Fort ~ St. Vrain Nuclear Generating Station for the period of January

'through June, 1987. . 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 sectiv,3 of our Te:hnical Specifications and 40CFR190, subpart B.

The following tables with a supplemental information section are

-included with this report:

Table Description 1A Gaseous Effluents - Summat+cu of All Releases

,. 1C Gaseous Effluents - Ground L? vel Releases ,

E t

.2A Liquid Effluents - Summatit, of All Releases l

'2B Liquid Effluents ';

D .3 Solid Waste and Irradiated Fuel Shipments

- 4A Hourly Meteorological'l Data i

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 i 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 i Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, I

. Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. J L 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.

Sample activities that are less than the detection capabilities of I ld our. equipment are entered in this report using the value resulting l

, 'from the calculation of the lower limit of dete'ction (LLD) or minimum detectable activity (MDA). This results in reporting upper limit values that are in excess of true activities, g n.

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P-87297 August 24, 1987 I

I The lower limit of detection (LLD), for the purpose of this report, i is defined as the smallest concentration of radioactive material in a I sample that will yield a net count, above system background, that '

will be detected with a 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 i follows: '

Liquid I l

Principle Gamma Emitters 5.00E-07 pCf/mi f

Dissolved Noble Gases 1.00E-05 pCi/ml '

i Tritium 1.00E-05 pCi/mi !

Iodine-131 1.00C-06 pCi/mi Gross Alpha 1.00E-07 pCi/mi Strontium-89,90_(Composite) E.00E-08 pCi/ml 1 i

Gaseous l l

Principle Gamma Emitters 1.00E-04 pC1/cc i (Gas) l-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 pCi/cc l

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 LLO values and the total volume of the media. "Less-than" values were not included ;

in the total values for the pathway.

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I The percent of Technical Specification limit on Table 1A is blank in some cases 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 are not reported 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 spe-ifications.

4 There has been some confusion in the past as to the total volume of ]

water used for dilution of a radioactive liquid effluent. All I average diluted concentrations are based on the activity at the unrest.icted area. Although this effluent will 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 section 4a) of the Supplemental Information Section.

An unanalyzed liquid waste release was made from the liquid waste system monitor tank on February 7, 1987. Based upon a maximam sample ,

activity (year to date) of all samples analyzed in 1987 and a total release volume of 6,458 liters, the total activity released was calculated to be 9.75E-01 curies. The cooling tower blowdown sample taken during the release substantiates the fact that the Maximum Permissible Concentration had not been exceeded. A complete description of the event can be found in Licensee Event Report 87-004 submitted to the Nuclear Regulatory Commission via Public Service Company letter P-87112, An unplanned gaseous release occurred on May 29, 1987, as a result of "A" Circul: tor seal leaking when the circulator was shutdown for maintenance. This activity was detected and recorded by the particulate monitor RT-73437-2 which increased from 160 to 400 counts per minute (cpm). Based on the exhaust flow rate, monitor sensitivity, and the release duration the total activity released was calculated to be 2.84E-05 curies, i

)

P-87297 August 24, 1987 An unplanned gaseous release occurred on June 5, 1987, while returning "B" Helium Circulator to service. A small arount of radioactive gas was released to the atmosphere through the reactor "j plant exhaust stack. The unplanned release was a result of activity in the circulator process piping, which was released as the circulator was being returned to service. This activity was detected and recorded by the particulate monitor RT-73437-2 which showed an '

increase of 100 counts per minute. Based on the exhaust flow rate, monitor sensitivity, and the release duration the total activity released was calculated to be 1.17E-05 curies, A description of the event can be found in the Non-Emergency Event Report 87-026 submitted to the Nuclear Regulatory Commission, and recorded in Public Service Company memo, PPS-87-2196, to the Plant Operations Review Committee.

Three unplanned gaseous releases occurred on June 12, 1987, when the System 46 head tanks were inadvertently vented to the reactor plant exhaust stack instead of the gas waste system. The activity was detected and recorded by the particulate monitor RT-73437-2 which increased from 200 to 375 counts per minute on the first release, 350 to 600 counts per minute on the second release, and 353 to 860 counts per minute on the third release. Based on the exhaust flow rate, monitor sensitivity, and the release duration for each release the total activity released was calculated to be 1.19E-04 curies. A description of the event can be found in the Non-Emergency Event Report 87-029 submitted to the Nuclear Regulatory Commission, and a recorded in Public Service Company memo, PPS-87-2493, to the Plant i Operations Review Committee, i There were no radioactive solid waste shipments made during this report period.

There were no major changes to the radioactive waste systems during ;

i this report.

There were no changes to the Process Control Program (SUSMAP-3),

Issue 2, effective date November 13, 1984, during this report period. J Ur.c change was made to the Offsite Dose Calculation Manual, SUSMAP-2.

Issue 14, effective date January 21, 1987, which 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 iodine cartridge. Also, a note was moved 3 to the beginning of a step and all references to the word 4

" instrument" were changed to " monitor." i r i

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P-87297 August 24, 1987 I

We are currently in the third quarter of a plan to evaluate an on line liquid beta monitor installed in the plant via Changa Notice 1 2313. The details during the third quarter were submitted to the j Nuclear Regulatory Commission via Public Service Company letter P-87237.

During this reporting period, three stack monitors and one recorder l for the Reactor Building ventilation exhaust system were inoperable

. for a period of greater than 30 days. Monitor RT-4803 was held over from the prior Semi-Annual Radioactive Ef fluent Release Report out of I service when it was discovered to be overdue for calibration. The holdup was 73 days and 17 hours1.967593e-4 days 0.00472 hours 2.810847e-5 weeks 6.4685e-6 months on SP 86501280 waiting for a i decision on safety relatedness. Monitor RT-6314-2 was out of service for 52 days and 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months on SSR 87503851 waiting for repair parts.

Monitor RT-7325-1 was declared out of service on January 16, 1986, for modifications to the module. The holdup was 44 days and 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months i on CN-1997 and CWP 86-0210 wait'ing for test equipment to test the l module. Recorder RR-93256 which supports RT-7324-1 and RT-7314-2 was ;

out of service.for 168 days and 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months on SSR 86514382. ' The holdup l was due to awaiting parts to fix the recorder and the performance of the Post Maintenance Test (PMT) that had not been done.

Per our Technical Specifications the Radioactive Effluent Release Report, submitted within 60 days af ter 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, q

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P-87297 August 24, 1987 1

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 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 particulate 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 particulate with half-lives greater than eight days, less than or equal to 1500 millirems per year to any organ.

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I The dose or dose commitment to a memt' of the public from I radioactive material in liquid effluent released to unrestricted l 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,

b. During any calendar year to less than or squel to 3 millirems to the total body and to less than or equal I to 10 millirems to any organ. l l

2. Maximum _ Permissible Concentrations i

All MPC values ured in determining allowable release rates from j the gas waste holdup system and the liquid waste system are those 1 listed in Table II, Columns 1 and 2 respectively, of Appendix B .

to 10CFR20. In addition, for the MPC of dissolved noble gases in l liquid effluent, the value of 2.00E-04 microcuries per milliliter {

was used.

3. Average Energy

)

1 The average energy (E-BAR) of the radionuclides mixture in release I 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 heldup 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 release header, control functions tre initiated wnich 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 1 approximately 32,000 cubic feet per minute. The full flow of this is directed through high efficiency particulate filters (HEPA) and activated charcoal beds prior to the release to the environment.

I P-87297 - August 26, 1987 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. Iodine For gaseous iodine, the Reactor Building exhavet ventilation is monitored and recorded on a continuous bases. The iodine cartridges used in these monitors are removed from service after one week and quantitatively . analyzed on a 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 iodine discussed in b. above, a particulate filter is removed and analyzed each week. Gross beta analyses are performed to identify and quantify any radionuclides. l 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. sfguid_ 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 plant site Outfall from th) Goosequill Pond rasches the South Platte River, both rivers converge approximately one and one half miles from the plant site. The averaga str.asm flow reported in section ba. of this supplemental report is a summation of both rivers and was calculated from data provided by the State of Colorado, Division of Water Resources in Greeley, Colorado.

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P-87297 August 24, 1987 l

l 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 radionuclides is tritium.

This discharge path utilizes a continuous sampler and an aliquot of this composite sampler is taken three times per {

weet 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 i Building Sump is made directly to the unrestricted area with j no dilution. 1

-I Tha batch release mode is used on the Reactor Building Sump l effluent and the liquid waste ' processing system. The l Reactor Building Sump can hold several hundred thousand '

gallons of waste water from various sources which could be contaminated. The liquid waste system consists of two 2000 j gallon receivers, one 2000 gallon monitoring tank and f associated filters and demineralizers. This system is j 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 j include gross beta, gross alpha, tritium and. gamma spectral l analyses. The result of these analyses and other analyses 4 as dictated by the gross beta results are used to determine the maximum release rate from the site. The liquid effluent is diluted with cooling tower blowdown which flows at a ;

minimum of 1100 gallons per minute. The resulting mixture 4 is sampled during the release period to confirm compliance l with regulatory limits. l The liquid effluent from the batch release mode is monitored !

continuously by redundant gamma activity monitors.

l l All tank level indicating devices, flow monitoring and ;

recording devices, and radiation monitor equipment are i calibrated and maintained at scheduled intervals in l accordance with established procedures. l l

Composite samples from the batch releases, and continuous l; releases are analy7ed monthly for Sr-89, Sr-90 and S-35.

All sample results are conservatively decay-corrected to the j start of the composite period. l l

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P-87297 August 24, 1987

v. Overall Error

~

d The overall error associated with determining the total act4 vity released from the site for both gaseous and liquid 1 effluent.is estimated to be 17.3 percent. This value is the I 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. ,

I

5. Batch Releases I

a. Liquid l Number of Batch Releases l 76 l l l I l Total Time Period For l l l Batch Releaset i 1.03E+03 HOURS 1 l l I l Maximum Time Period For l l la Batch Release l 4.38E+01 HOURS l

( l l l Average Time Period For l l la Batch Release i 1.35E+01 HOURS l I I I l Minimum Time Period For l l la Batch Release l 4.35E+00 HOURS l l .

I I l Average Stream Flow During l l l Period of Release of l l lEf fluent Into a Flowing l l l Stream l 7.16E+05 GPM* l l l l

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I P 87297 August 24,.1987

b. Gaseous .;

l Number of Batch Releases l 166 l L i I I l l70tal Time Period For l l l Batch Releases l 5.95E+02 HOURS l 1 1 I I l Maximum Time Period For l l la Batch Release l .1.49E+01 HOURS l l l l l Average Time Period For l .I ja Batch Release l 3.59E+00 HOURS l l I l l

-l Minimum Time Period For l l la Batch Release l 7.00E-01 HOURS I l l __ l l

6. b

'l A_bnormal Releases i

a. Liquid 1

l Number of Releases l 1.00E+00' l l 1 l l 1 l l Total Activity Released l 9.75E-01 CURIES l l l l. l l i

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b. Gaseous l .lNumber of Releases l 5.00E+00 l l l l '.

[ Total Activity Released l 1.59E-04 CURIES l I I I

1 TABLE 1A EFFLUENT AN0' WASTE DISPOSAL SEMIANNUAL REPORT 1987 i ,

GASEOUS EFFLUENT-SUMMATION OF ALL RELEASES l Unit l Quarter l Quarter l Est. Total l

[ l 1 l 2 l Errer. % l l l l l l A. Fission.and activation products

1. Total release l Ci l 0.00E+00 l 6.24E+01 l 1.70E+01 J l i l l l

2. Average release rate for luCi/secl 0.00E+00 l 7.94E+00 l.

period l l l l 1 l l l

3. Percent of technical l % l 0.00E+00 l .6.41E-01 l specification limit l l l _l B. Iodine I i

1. Total iodine-131 l C.i l <4.64E-06 l <4.69E-06 l l '

l l l ~l l

2. Average release rate for luci/secl <5.96E-07 l <5.96E-0/ l period l l l l l l l l s

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

1. Particulate with half-lives l Ci l <1.89E-07 l <2.34E-07 l l

> 8 days l l l l l l l l _.I

2. Average release rate for luc 1/secl <2.44E-08 l <2.98E-08 l period l l l l l l 1 l

3. Percent of technical l % l l l specification limit l l l l l l _I I ,

4. Gross alpha radioactivity l C1 l <4.13E-08 l <6.23E-08 l ;

I l l l l

D. Tritium i

1. Total release l Ci l 1.29E-02 l 1.30E+00 l 1.70E+01 l l l l-. I ~I l

2. Average release rate for luc 1/secl 1.66E-03 l 1.65E-01 l j period l l l l q

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3. Percent of tech. spec. 1 % l 2.77E-05 l 2.75E-03 l 1 limit l 1 l l l l

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i-TABLE IC- 'l l l L , EFFLlENT AND WASTE DISPO^AL SEMIANNUAL REPORT (1987) , j GASE0US EFFLUENTS--GROUND-LEVEL RELEASE CONTINU0US MODE BATCH MODE lReleesed l Unit I Quarter ! Q arter l Quarter 1 l Quarter 2 l '

l l 1 l L I I Fission gases

]

l kry,nton-85 l C1 l l l <1.01E+01 l <1.09E+01 1 I -

L 1 l l_ l. l. .

Ikrypton-8bm l Ci l l l l 2.16E-01 l 1 I I I I l I I I

I krypt.on-87 l Ci l l ,' <7.29E-02 l 1.16E-02 l l -- - I I I I I l

.lkrypton-88 l Ci l l l71.11E-01 l 1.52E-01 l l 1 I I I I I lxenon-133 l Ci l l l <1.06E-01 'l 5.97E+01 l l 1 I I I I I lxenon-135 I Ci l l l <3.05E-02 l 1.60E+00 l l l l l I l I lxenon-135m I ci l l l l _l )

'l l I I l l 1 )

l xenon-138 l Ci l l l <3.59E-01 l <3.89E-01 l l l I I I l- l l l lxenon-133m l C1 I l l <2.45E-01 1 6.48E-01 l 1 I .I I l l- I lxenon-131m l C1 l l l l l 1 I I I I l i I I ITotal for period

I Ci l l l 0.00E+00 l 6.23E+01 l l l l l l l ._I Iodinet i l

4 l iodine-131 l Ci j l l <1.74E-09 I <1.89E-09 l l l l l l l l liodine-133 I C1 l l l l l l -

l l l l L i l iodine-135 l Ci l l l l l l l l l 1 I l ll Total for period I Ci l l l 0.00E+00 1 0.00E+00 l l l I I I I I 1

l Total values do not include "<" data l l

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Page 2 TABLE 1C (Continued)

I Particulate l

istrontium-89 l Ci l l l l l Y l- 1 I I I I .I

-l strontium l Ci l l l l l ?

l. l l l- I l .- l -

l cesium-134 l Ci l l l l l l l l l l l l l ce s i un.-137 . l Ci ) l l l l l l .

I I I I I I

'i barium-lanthanum-140 l Ci l l l l l l 1 1 I I i 1 i, i

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Total values do not include "<" data

TABLE 2A EFFLUENT AND WASTE DISPOSAL SEMIANNVAL REPORT 1987 l

l LIQUID EFFLVENT-SUMMATION OF ALL RELEASES .

! Nuclide l Onits l Quarter l Quarter j Est. Total'l ;

l l l 1 1 2 l Error. % l l I I I I l l ;

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i A. Fission and activation products

1. Total release lCs-137 l C1 l 0.00E+00 l 1.18E-06 l 1.70E+01 l ) '

l l l ! I l

2. Average diluted luci/ml l 0.00E+00 1 2.16E-12 l concentration [ l l l l I I l 1 3 .' Percent of I % l l 1.08E-05 l applicable limit l l l .l L

1. Total release lCo-60 l Ci l 0.00E+00 1 0.00E+00 1 1.70E+01 l l l l l l. I

2. Average diluted luCi/mi l 0.00E+00 l 0.00E+00 l concentration l l l l l l l l

3. Percent of l % , I l applicable limit l., I l l ,

B. Tritium t

1. Total release lH-3 l Ci l 4.68E+00 l 2.93E+01 l 1.70E+01 l !

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2. Average diluted luCi/mi l 8.67E-06 l 5.37[-05 i concentration l l l l l l l l 1

3. Percent of l % l 2.89E-01 l 1.79E+00 l !

applicable limit j l l l ]

l C. Dissolved and entrained ]

gaset J I

1. Total release lXe-133 l Ci l 0.00E+00 l 1.79E-05 l 1.70E+01 l l l I l l l l l

2. Average diluted luci/ml l 0.00E+00 l 3.28E-11 l 1 concentration l l l l l l l l

3. Percent of l % l l 1.64E-05 l applicable limit l_ l l l

0. Gross alpha radioactivity

1. To+.al release l Ci j<8.90E-05 1 2.94E-05 l 1.70E+01 l l l l 1 I l

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. Page 2 TABLE 2A (Continued)

E. Volume of warte released (prior to dflution) l Liters l 1.50E+07 l 1.67E+07 l 1.00E+01 l l l l l l F. Volume of dilution water used during release ILiters l 5.40E+08 l 5.46E+08 l 1.00E+01 l

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TABLE 2B

, EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1987) ,

LIQUID EFFLUENTS 1

CONTINU0US MODE BATCH MODE lNuclides Released l Unii, l Quarter 1 l Quarter Z l Quarter 1 l Quarter 2 l l l l l l l l strontium-89 l Ci l <2.16E-05 i <2.08E-0T)l <6.14E-06 l <6.71E l l l l - 1 l l strontium-90 l Ci l <2.26E-05 l <2.19E l D El'<6.39E-96 I <7.03'E l' I I I I l~ l l cEium-134 [ Ci l <8.31E-04 I <8.92E-04"l'72.18E"-04 l <2.7CE-04 l l __ l i -

l . _ _ 1 I _. I l cesium-137 I C1 l <8.24E-04 l <8.M E-04 i <C.17E-04 l <2.69E-04 1 l .

I I l ! l .__. I l iodine-131 l Ci l <6.41E-04 l <6.89E-04 1 <1.69E-04 l <2.14E-04 l I _ _ l I I l_ l I

~

l cobalt-58 ! Ci I <6.99E-04 l <f.51E-04 l <1.04E'-04 l <2.34E-04 l l -

._. i _I - - _.l___ l __._ I -__I I cobalt-60 1 C1 l <8.76E-04 l <9.41E-04 l <2.31E-04 l <2.94E-b4 l l _. I - _ _l l l 1. -_.

I l iron-59 I Ci l <1.35E-03 1 <1.45E-03 i <3.55E-04 l <4.53E-04 l l l l l l l l l zinc-65 l Ci l <1.73E-03 l <1.86E-03 l <4.b5E-04 l <5.80E-04 l ,

l l l l_ l --l l I manganese-54 l Ci l <6.63E-04 i <7.12E-04 l <1.74E-04 l <2.22E-04 l l_ _I I I I I _. I ;

l chromium-51 l Ci l l l l l !

I I I I _I I I

_ l l zi rconi t.m-ni obi um-95 l Ci l l l l l j l I _._._ ____. I l _l l J l molybdenum-99 l Ci l <5.24E-03 l <5.62E-03 l <1.38E-03 l <1.75E-03 l ] '

I I I ; I I I l technetium-99m I Ci l l j l l j L l___ l I l_ l I j l carium-lanthanum-140 l Ci l l l l l l .. _ l l I I l l l cerium-141 l C1 1 <1.18E-03 l <1.27E-03 l <3.11E-04 l <3.96E-04 l l l 1 I i l i I tritium l Ci l 6.08E-02 1 1.25E-01 1 4.62E+C0 l 2.92E+01 l

l. I I 1. I I I l sulfur-35 } Ci l <2.77E-03 l <2.07E-03 l <6.52E-04 l <9.27E-04 l l l l l l l l

~

l Tetal for period (above)*l Cd l 6.08E-02 l 1.25E-01 1 4.62E+00 I .2.92E+01 l l 1 l l I I I I Total values do not include "<" data !

c .

' i

.l I 1 f a i

< Page 2 TABLE 28 (Continued) s

+ l 4

l

' 't xenon-133 l C1 1 <2.29E-03 l <2.46E-03 l <6.02E-04 l <7.61E-04 l j l I l l I '1 ~

f [:

xenon-133 l Ci l <5.93E-04 l <6.36E-04 l <1.57E-04 ll <1.99E 0 1

'\ . l .I \ _. I l l

.1 i

,i s r

f

c. i

'l i

l 1

.s i,.

i t

I i

i l

l i

l l

Total values do not include "<" data l i

VABLE 3

, EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1986) j i

SOLIO WASTE AND IRRADIATED FUEL SHIPMEiiTS 1

A. SOLIO WASTE SHIPPE0 0FFSITE FOR BURIAL OR DISPOSAL (Not irradiated fu?.) i c

'} 1. Type of waste l Unit l 6-month I Est. Total (

l' i 'l Period i Error % 'l 1 l l l l l I I I I i !

I a. Spent resins,-filter l m2 l 0.00E+00 l l 1 l sludges', evaporator l Ci l 0.00E+00 l l i

i bottoms, etc. l l l 0.00E+00 l l l I l I ;

I I I I l' l I

b. Dry compressible waste, I m' l 0.00E+00 l l j l contaminated equip. etc. I Ci l 0.00E+00 l 0.00E+00 l- j l' I I l: I i l I I I I I c. Irradiated components, I m' l 0.00E+00 l l ,

I control rods, etc. 1 Ci l 0.00E+00 1 0.00E+00 l 1 i L 4 I l' I I I I l ;

I d. Other(describr) l m' i 0.CDE+00 l l !

{ l Ci 1 0.00E+00 1 0.00E+00 I l l_ --_ .I __ ) &___ m I i

' I,

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

% l l

% l l 1

b. I % l- I l I % l j 1 % l l

c. , _ , I % ., 1 I

_ L *4 I

1 4 I i l

d. I % l l

% 1 L

% l l

3. Sclid Waste Disposition Number of Shipments Mode of Transportation Destination 0

1 B. IRRADIATED FUEL SHIPMENTS (Disposition) '

, Number of Shipments Mode of Transportation Destination ;

j 0

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4 h Public Service ~ Patim...
l August 24, 1987 Fort St. Vrain Unit No. 1 P-87297 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 Docket No. 50-267
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 Staticn.
This report covers the period January 1, 1987 through June 30, 1987, 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, Ah ..
R. O. Williams, Jr.
Vice President, Nuclear Operations Fort St. Vrain Nuclear Generating Station R0W:00M/skd Attachments ZG49 Ilt
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I l
P-87297 August 24, 1987 l l
l cc w/ attachments: I i
Regional Administrator, Region IV )
V. S. Nuclear Regulatory Commission j Attn: Mr. J. E. Gagliardo, Chief l Reactor Projects Branch l 611 Ryan Plaza Drive, Suite 1000 l Arlington, Texas 76011 i l
Dottie Sherman, ANI Library i American Nuclear Insurers .
The Exchange Suite, 270 Farmington Avenue Farmington, Connecticut 06032 ,
l Mr. Albert J. Hazie, Director l Radiation Control Division Colorado Department of Health l 4210 E. lith Ave.
Denver,'C0 80220 Mr. Mike Dolphin, Site Manager General Atomic Company P.O. Box 426 Platteville, CO 80651 ,
1' 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
a P-87297 A .st 24, 1987 Or. James E. Johnson Radiology & Radiatien Biology Dept.
Colorado State University Fort Collins, CO 80523 Mr.. Robert Farrell Senior NRC Resident Inspector Fort St. Vrafn Statit-i 1
_ _ _ _ - _ _ - - _ _ _ _ -

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