Text

Semiannual Radioactive Effluent Release Rept Jul-Dec 1988
	ML20235V589
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/31/1988
From:	Robert Williams; PUBLIC SERVICE CO. OF COLORADO
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	P-89073, NUDOCS 8903100244
	Download: ML20235V589 (52)
	v • d • e

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SEMI-ANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT July - December 1988 Public Service Company of Colorado Fort St. Vrain Nuclear Generating Station February 1989 8903100244 881231 PDR ADOCK 05000267 PDC R

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s P-89073 February 28, 1989 This report summarizes the radiological effluent released from the Fort St..Vrain Nuclear Generating Station for the period of July through December, 1988. -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 1A 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 1B (of Regulatory Guide 1.21) has been omitted' from this report because all of our gaseous effluents are assumed to be ground-level releases as opposed to being 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, 2n-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 cf the above nuclides, are analyzed and included in this TP "J o rt.

Sample activities that are less than the detection capabilities of our equipment are entered in this report using the value resulting from the calculation of the lower limit of detection (LLD) or minimum detectable activity (MDA). This results in reporting upper limit values that are in excess of true activities.

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P-89073 February 28, 1989 The lower limit of detection (LLD), for the purpose of this report, is defined as the smallest concentration of radioactive material in a sample that will yield a net count, above the system background, that will be detected with a 95% probability of being correct and only a 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 pC1/ml Dissolved Noble Gases 1.00E-05 pCi/mi Tritium 1.00E-05 pC1/mi Iodine-131 1.00E-06 pCi/ml Gross Alpha 1.00E-07 pCi/mi Strontium-89,90(Composite) 5.00E-08 pCi/ml 1

Gaseous Principle Gamma Emitters 1.00E-04 pCi/cc (Gas)

Principle Gamma Emitters 1.00E-11 pCi/cc (Particulate) i Tritium (Gas) 1.00E-06 pCi/cc Iodine-131 (Charcoal) 1.00E-12 pCi/cc GrossAlpha(Particulate) 1.00E-11 pCi/cc Strontium-89, 90 (Particulate) 1.00E-11 pCi/cc Gross-Beta (Particulate) 1.00E-11 pC1/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 I

the total volume of the media.

The "less-than" values were not I

included in the total values for the pathway.

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P-89073 February 28, 1989 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 1C, the continuous release mode values are not reported because this release pathway is the same as the batch mode. All other blanks on Tables IC 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 in the past 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 an'd South Platte River) which converge approximately one and one half miles downstream of the plant, no further dilutions were assumed. Additional discussion on river flow is contained in section 4a) of the Supplemental Information Section.

One incident of tritium introduction into System 42 occurred between July 13, 1988 and July 18, 1988 after which the levels returned to normal. During this event,1.28 E-01 Ci of tritium was released when 72,000 gallons of System 42 service water overflowed to the blowdown stream.

In an effort to remove chemical contamination in the condensate storage tank, part of the stored water was drained to System 42 prior to refilling with makeup water. This was reported as an abnormal release in the third quarter. Similar events in earlier report periods had been traced to a malfunction of pressure relief valve (V-31367).

The drain pipe from this relief valve was rerouted from the service water return sump to the turbine building sump.

This work was begun on October 31, 1988 under Change Notice #2812 and Controlled Work Procedure #88-0224 and completed on November 4, 1988.

No similar events have happened since that time.

1 In another effort to remove chemical contamination in the condensate storage tank, 210,000 gallons cf water was released to the South Yard Drain on July 26, 1988 The total tritium activity released was 3.70 E-01 Ci.

l During this reporting period, there were a total of 15 times when primary coolant was intentionally released through the purification train without liquid nitrogen in the low temperature filter absorber.

The total activity released was 1.84 E-02 Ci of 133 Xe and 6.95 E-03 Ci of tritium.

These were reported as abnormal releases during both quarters.

There were no unplanned radioactive liquid waste releases made during this reporting period.

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P P-89073 February 28, 1989 There were no unplanned radioactive cas waste releases made during this reporting period.

The following discussion correlates specific points mentioned in Fort St. Vrain Technical Specification 7.5.1.e to the contents of the WASTETRAK generated. Regulatory Guide 1.21 Radioactive Waste Report for the Semiannual Effluent Report.

Total volumes and curie quantities are given for each waste type-in Table 3.

The curie content of each container was estimated by the WASTETRAK computer code. WASTETRAK calculated the concentration of gamma emitting nuclides from the measured radiation level on each container and, after applying.the appropriate scaling factor to obtain the concentration of difficult to measure radionuclides summed the concentrations and calculated a total curie content for the container.

Principal radionuclides were estimated by the WASTETRAK computer code based on scaling factors and nuclide distribution determined from direct and representative samples as part of the 10CFR61 Waste Classification program performed in 1986.

All waste disposed of was either Process Waste or Dry Active Waste.

No Irradiated Components were shipped for disposal in 1987.

All waste disposed of was Low Specific Activity Waste. Absorbents were added to some Process Waste but no solidification agents or absorbents were employed in processing Dry Active Waste disposed of in 1987.

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

There were no changes to the Process Control Program, SUSMAP-3, Issue 2, effective date November 13, 1984, during this reporting period.

There were no changes to the Offsite Dose Calculation Manual, SUSMAP-2, Issue 16, effective date May 3,1988 during this reporting period.

P-89073 February 28, 1989 I

There were five effluent release monitors out of service for more than 30 days during this reporting period.

RIS-4801, RIS-4802 and RIS-4803, the backup stack monitors, were out of service because of excessive heat degradation. The lack of spare parts delayed their calibration and return to operation.

RIS-7324-2, a noble gas monitor, was out of service because the pump could not supply sufficient flow.

The time necessary to rebuild the pump delayed return to operation.

RIS-9301, the primary coolant monitor was out of service because of detector problems and its return to service was delayed by the lack of the spare parts necessary to rebuild it.

Radiation doses resulting from the release of radioactive liquid and gaseous effluents from Fort St. Vrain during 1988 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.

Ooses are calculated for a hypothetical " maximum" individual present at all times of the year at the Exclusion Area Boundary (EAB) in the sector in which all of the activity is calculated to have been released.

Liquid - 10CFR50 Whole Body 4.00E-01 mrem Bone 1.73E-01 mrem Liver 4.81E-01 mrem Thyroid 2.49E-01 mrem Kidney 3.26E-01 mrem Lung 2.72E-01 mrem GI 2.50E-01 mrem

P-89073 February 28, 1989 Gaseous - 10CFR50 Noble Gas Gamma 5.29E-01 mrad Beta 1.31E+00 mrad Iodine, Particulate, Tritium Adult Whole Body 1.08E-01 mrem Organ (maximum) 1.08E-01 mrem Bone 0.00E+01 mrem Teen Whole Body 1.24E-01 mrem Organ (maximum) 1.24E-01 mrem Bone 0.00E+01 mrem Child Whole Body 1.78E-01 mrem Organ (maximum) 1.78E-01 mrem Bone 0.00E+01 mrem Infant Whole Body 1.39E-01 mrem Organ (maximum) 1.39E-01 mrem Bone 0.00E+01 mrem Gaseous - 10CFR20 Iodine, Particulate, Tritium 2.78E+02 mrem All doses are within acceptable limits in accordance with 10CFR20 and 10CFR50.

It was felt that use of actual dilution factors was more accurate than the annual average X/Q of 1.37E-06 s/m3 as previously reported in the Final Safety Analysis Report. Beginning with 1987, dilution factors for actual periods of release were used to calculate doses from gaseous effluent releases.

The doses from gaseous releases are calculated including contribution from direct radiation. A review of Termoluminescense Oosimeter data collected in 1986 has confirmed that no measurable direct radiation exposure is attributable to Fort St. Vrain.

P-89073 February 28, 1989 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:

1)

Residents living within the site boundary are at home during 50% of gaseous effluent releases.

2)

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.

3)

All other assumptions of SUSMAP-2 remain valid.

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 less than 3.06E+00, The 29.78 key iodine-129 peak is below the minimum detectable energy of our detectors. We assume that no iodine-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 i

value of gross alpha is listed in Table 2A Section D.

An annual land use census is performed as part of the Fort St. Vrain Radiological Environmental Monitoring Program.

Changes made to environmental sampling locations as a result of the annual land use census are reported in the annual Radiological Environmental Monitoring Program report.

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February 28, 1989 EFFLUENT AND WASTE DISPOSAL SEMIANNUAL 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 i

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

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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

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the unrestricted area is limited to:

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5 millirads gamma and 10 millirads beta during any i

calendar quarter, and, j

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 iimited to-1 a.

7.5 millirems to any organ during any calendar quarter,

and, i

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 grester than eight days, less than or equal I

to 1500 millirems per year to any organ.

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P-89073 February 28, 1989 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, 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 Maximum Permissible Concentration (MPC) values 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.

3. ' Average Energy The average energy (E-BAR) of the radionuclida 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 holdup system are performed after sampling and analyses for noble gases and tritium.

T'ese analytical results are nsed along with atmospheric dilution factors to determine the allowable release rate. Gas is released on a contiruous 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 are initiated which divert the gas to the gas waste holdup system.

s, P-89073 February 28, 1989 All radioactive gases are released to the Reactor. Building exhaust ventilation system which has a flow rate 'of approximately 30,000- 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 Downstream of the acdyned 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 exhaust ventilation is monitored and recorded on a continuous basis. The iodine cartridges used in these monitors are removed after one week of service and quantitatively analyzed on a

gamma spectroscopy system.

The quantity of radiciodine released ~

during that period is calculated based on.the integrated

. flow during the co'llection period.

c.

Particulate As in -the case of the iodine discussed in b. above, a particulate filter is removed and analyzed each week. Gross beta analysis as well as gamma spectral analysis is performed to identify and quantify any radionuclides.

The quantities 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 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 the Goosequill Pond reaches the South 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 received and tabulated from data provided by the Colorado Department of Natural Resources in Greeley, Colorado.

U P-89073.

' February 28, 1989-

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 p

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 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 sampla 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 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 gallon receivers, one 2000 gallon. monitoring tank and associated filters and demineralizers.

This system Jis

' 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 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 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 mair,tained 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 i

start of the composite period.

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P-89073 February 28, 1989 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

117 l

j.

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l Total Time Period for l

l l Batch Releases l

1.13E+03 HOURS l

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l l Maximum Time Period for l

l la Batch Release l

8.07E+01 HOURS l

l l

l l Average Time Period for l

l la Batch Release l

9.65E-01 HOURS l

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l l Minimum Time Period for l

l la Batch Release l

3.33E+00 HOURS l

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l l Average Stream flow During l

l l Period of Effluent Release l 2.40E+05 GALLONS PER l lInto a Flowing Stream l

MINUTE 1

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P-89073 February 28, 1989 b.

Gaseous l Number.of Batch Releases l

139 l

l l

l l Total Time Period for l

l la Batch Release l

4.42E+02 HOURS l

l l

l l Maximum Time Period for I

l la Batch Release l

1.30E+01 HOURS l

l l

l l Average Time Period for l

l la Batch Release l

3.18E+00 HOURS l

l l

l l Minimum Time Period for l

l la Batch Release l

1.10E+00 HOURS

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6.

Abnormal Releases a.

Liquid l Number of Releases l

2 l

l 1

l l Total Activity Released l

4.98E-01 CURIES l

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I b.

Gaseous l Number of Releases l

15 l

l 1

I l Total Activity Released l

2.54E-02 CURIES I

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.O
TABLE 1A EFFLUENT AND dASTE DISPOSAL SEMIANNUAL REPORT 1988 GASE0US EFFLUENT-SUMMATION OF ALL RELEASES l Unit l Quarter l Quarter l Est. Total l l
l 3
l 4
l Error. %
l l
l 1
l l
A.
Fission and activation products 1.
Total release l
Ci l
6.08E+01 l 3.62E-01 l 1.73E+01 l
l l
l I
I 2.
Average release rate for juCi/seci 7.65E+00 l 4.55E-02 l period l
l l
l 1
l 1
l 3.
Percent of technical l
1 1.40E+00 1 3.64E-03 l specification limit l
l l
l B.
Iodine 1.
Total iodine-131 l
Ci l <3.21E-06 l <3.21E-06 l l
l l
1 I
i 2.
Average release rate for luc 1/ sect <4.04E-07 l <4.04E-07 l period l
l l
l 1
I l
l 3.
Percent of technical l
l l
l specification limit l
l l
l C.
Particulate 1.
Particulate with half-lives l
Ci l <2.21E-07 l <1.53E-07 l l.
> 8 days l
l l
l l
l l
l l
2.
Average release rate for luCi/secl <2.78E-08 l <1.93E-08 l period l
l l
l l
l l
I 3.
Percent of technical I
i l
i specification limit l
l l
l l
1 I
I 4.
Gross alpha radioactivity l
Ci l <3.01E-08 l <3.43E-08 l 1
I I
I D.
Tritium 1.
Total release l
Ci l
6.40E-01 1 1.20E-01 l 1.73E+01 l
l l
l l
l 2.
Average release rate for luc 1/seci 8.05E-02 l 1.51E-02 l i
period l
l l
l l
l l
l 3.
Percent of tech. spec.
l l
1.34E-03 l 2.52E-04 l limit l
l l
l l
l
\\
TABLE IC 1
EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1988)
GASE0US EFFLUENTS--GROUND-LEVEL RELEASE CONTINU0US MODE BATCH MODE l Released l Unit l l
l Quarter 3 l Quarter 4 l l
1 l
l l
l 1
Fission gases j krypton-85 l
C1 l
l l
5.87E-01 l 3.62E-01 l i
1.
I I
I l
l l krypton-85m l
Ci l
l l
1.83E-02 l l
l l
1 I
I I
l l krypton-87 l
Ci l
l l <2.25E-02 l <7.75E-02 l l
l-l l
l l
I l krypton-88 l
Ci l
l l
5.97E-03 l <1.15E-01 l l
1 l
l l
l l
1 xenon-133 l
Ci l
l l
5.79E+01 l <1.08E-01 l l
l I
l l
l 1
'l lxenon-135 i
Ci l
l l
3.27E-01 l <3.27E-02 l 1
I I
I I
I I
[ xenon-135m l
C1 l
l l
l l
l-1 I
I l
1 I
l xenon-138 l
Ci l
l l <1.15E-01 l <7.38E-06 l l
l l
l 1
l l
lxenon-133m l
Ci l
l l
1.09E+00 l <5.29E-06 l l
1 I
l l
l l
lxenon-131m l
Ci l
l l
9.53E-01 l l
i l
l l
l l
l l
l Total for period
~~~~l Ci l~~~~
l l
6.09E+01 1 3.62E-01 l l
l 1
I l
l 1
Iodines liodine-131 l
C1 l
l l <4.45E-10 l <1.53E-09 l l
l l
l l
l 1
l iodine-133 l
Ci l
l l
l l
l l
i I
l l
1 liodine-135 l
C1 l
l l
l l
l l
l l
l 1
I l Total for period l
C1 l
l l
0.00E-01 l 0.00E-01 l l
1 1
I l
l 1
Total values do not include "<" data
i TABLEIC(Continued)
Particulate Istrontium-89 l
Ci l
l l
l
[
I I
l l
l l
'l l strontium-90 l
Ci l
l l
l l
1 I
I I
l l
l l cesium-134 l
Ci l
l l
l
[
l i
I I
I I
i
[ cesium-137 l Ci l
l l
l l
l l
l l
1 I
I lbarium-lanthanum-140 l
Ci l
l l
l l
l l
l l
1 I
I Total values do not include "<" data o
TABLE 2A EFFLUENT AND WASTE DISPOSAL. SEMIANNUAL REPORT.1988 1
LIQUID EFFLUENT-SUMMATION OF ALL RELEASES l
l Nuclide l Units l Quarter ! Quarter l Est. Total l l
l l
3 l
4 l
Error. %.
l l
l l
1 l
l A.
Fission and activation products 1.
Total release lCs-137 l
Ci l 1.04E-06 l 1.64E-04 l 1.73E+01 l
l 1
I l
l 1
2.
Average diluted luCi/ml l 1.89E-12 l 2.97E-10 l concentration l
l l
l l
l I
I 3.
Percent of l
l 9.45E-06 l 1.48E-03 l applicable limit l
l l
l 1.
Total release (Co-60 l
Ci 1 0.00E-01 l 0.00E-01 l l
1 1
I I
l l
2.
Average diluted luci/ml l 0.00E-01 1 0.00E-01 l concentration l
l l
l l
l l
l 3.
Percent of l
l l
l applicable limit l
l l
l B.
Tritium 1.
Total release lH-3 l
Ci l 3.97E+01 1 2.23E+01 l 1.73E+01 l
1 l
I I
I I
2.
Average diluted juC1/mi l 7.20E-05 1 4.04E-05 l concentration l
l l
l l
t l
1 3.
Percent of I
l 2.40E+00 l 1.35E+00 l applicable limit l
l l
l C.
Dissolved and entrained gases 1.
Total release lXe-133 l
Ci l 3.13E-03 l 0.00E-01 l 1.73E+01 l
l l
l l
1 l
2.
Average diluted luCi/mi l 5.67E-09 l 0.00E-01 1 concentration l
l l
l l
l l
l 3.
Percent of l
l 2.84E-03 l l
applicable limit l
l l
l s
l
TABLE 2A(Continued) 1 l
D.
Gross alpha radioactivity 1.
Total release-l Ci j 6.69E-04 l 1.10E-05 l 1.73E+01 l
l l
l 1
1 E.
Volume of waste released (prior.todilution)
ILiters l 1.74E+07 l 1.56E+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 1
TABLE 2B EFFLUENT AND WASTE DISPOSAL SEMIANNUAL REPORT (1988)
LIQUID EFFLUENTS CONTINUQUS MODE BATCH MODE lNuclides Released l Unit l Quarter 3 l Quarter 4 l Quarter 3 l Quarter 4 l i
l l
l.
l l
l l
\\
l strontium-89 i
C1 l <2.92E-05 l <2.39E-05 l <1.65E-05 l <5.89E-06 l 1
l 1
l l
l 1
l strontium-90 l
C1 l <3.04F-05 l <2.49E-05 l <1.72E-05 l <5.53E-06 l l-1 l
l l
l l
l cesium-134 l
Ci l <1.05E-03 l <9.23E-04 l <1.99E-04 l <2.02E-04 I I
I I
i l
1 I
l cesium-137 l
Ci i <1.00E-03 l <8.83E-04 l <1.78E-04 l <3.25E-04 l l
l 1
1 I
l l
l iodine-131 l
Ci j <7.98E-04 l <7.04E-04 1 <2.69E-05 l <1.54E-04 l 1
i I
l 1
I l
l cobalt-58 l
C1 l <9.43E-04 l <8.33E-04 l <1.80E-04 l <1.82E-04 1
'l l
l l.
l l
1 l
cobalt-60 l
Ci l <9.62E-04 l <8.49E-04 l <1.84E-04 l <1.85E-04 l l
l 1
l l
l l
l 1ron-59 l
Ci l <1.79E-03 l <1.58E-03 l <3.42E-04 l <3.46E-04 l 1
l l
1 1
I l
l zinc-65 l
Ci l <2.24E-03 1 <1.98E-03 l <4.28E-04 l <4.33E-04 I i
l l
l l
l l
l manganese-54 l
Ci l <8.97E-04 I <7.91E-04 l <1.71E-04 l <1.73E-04 l l
1 l
l l_
l 1
l chromium-51 l
Ci l
l l
l l
l 1
l l
l l
l l~
~.irconium-niobium-95 l
Ci l
l l
l l
l l
1 l
l l
l l
molybdenum-99 l
Ci l <6.88E-03 l <6.08E-03 l <1.32E-03 l <1.33E-03 l l
l l
l l
l l
l technetium-99m l
Ci l
l l
l l
l I
l l
l 1
I l
barium-lanthanum-140 l
Ci j
l l
l l
l 1
1 I
l l
l l
l I
cerium-141 l
Ci l <1.43E-03 l <1.26E-03 l <2.73E-04 1 <2.76E-04 1 4
I I
I l
l l
l l
tritium l
Ci l
1.34E+00 l 6.44E-02 l 3.84E+01 l 2.22E+01 l 1
I I
l l
1 I
l sulfur-35 l
Ci l <3.11E-03 1 <3.07E-03 l <1.01E-03 l <8.77E-04 l l
l l
1 l
l l
l Total for period (above)*j Ci l
1.34E+00 l 6.44E-02 l 3.84E+01 l 2.22E+0T 1 1
I l
l I
I l
l I
1 o Total values do not include "<" data
TABLE 2B(Continued)
I l
l l
Continuous Mode Batch Mode l
xenon-133 l
Ci l <2.89E-03 l <2.56E-03 1 <3.64E-03 l <5.59E-04 l l
1 I
I I
I I
(
l xenon-135 l
Ci l <7.63E-04 l <6.74E-04 l <1.46E-04 1 <1.47E-04 l l
l l
1 I
I I
I Trtal values do not include "<" data
TABLE 3 EFFLUENTANDMASTEDISPOSALSEMIANNUALREPORT(1988)
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 I
I I
I I
I I
I I
I I
l a.
Process waste, i.e.,
l m'
l 4.88E+00 l
l l
spent resins, filter l
Ci l
1.80E-01 l
7.55E-01 l I
sludges, evaporator l
l l
l l
bottoms, etc.
l l
l l
l 1
l l
I I
i l
I I
l b.
Dry active waste i.e.,
l m'
l 2.12E+00 l
l l
dry compressible waste, l
Ci l
1.30E-01 l
3.64E+00 l l
contaminated equip. etc.
l l
l l
I I
I I
I l
I I
I I
I i
l c.
Irradiated components, I
m' l
l l
l l
control rods, etc.
l Ci l
l l
l l
l I
I I
I I
I I
l d.
Other(describe) l m'
l l
l 1
l Ci l
l l
l l
l l
l 2.
Estimate of major nuclide composition (by type of waste)
Type of Waste l Isotope lContenti Curies l
Error %
l l
l
.I I
I a.
Process waste IH3 l 56.7% i 1.02E-01 l
1.06E+00 l
lC14 l 42.6% l 7.66E-02 l
1.06E+00 l
IN163 1
0.3% l 6.12E-04 l
1.49E+00 l
lFe55 1
0.2% l 3.14E-04 l
1.51E+00 l
lPu241 l
0.1% l 2.44E-04 l
1.49E+00 l
b.
Dry active waste lH3 l 86.2% l 1.12E-01 1
4.06E+00 l
IFe55 l 10.0% 1 1.30E-02 l
9.08E+00 l
1S35 l
2.0% l 2.64E-03 l
8.22E+00 I
lCs137 1 0.5% l 6.99E-04 1
3.86E+00 l
lAg110m 1 0.2% l 2.62E-04 1
4.24E+00 l
1 c.
Irradiated Compo-l l
l l
l nents l
l l
l l
d.
Other l
l l
l l
3.
Solid Waste Disposition Number of Shipments Mode of Transportation Destination 1
Highway Beatty, Nevada j
B.
IRRADIATED FUEL SHIPMENTS (Disposition)
Number of Shipments Mode of Transportation Destination
_.0__
p r
Qp Public Service *
,,,,,,,,a Company of Colorado P.O. Box 840 Denver, CO 80201- 0840 February 28, 1989 Fort St. Vrain R.O. WILLIAMS, JR.
SENIOn VICE PRESIDENT Unit No. 1 NUCLEAR OPE RATIONS P-89073
'U. S. Nuclear Regulatory Commission ATTN:
Document Control Desk
)
Washington, D.C.
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 Station.
This report covers the period July 1, 1988 through December 31, 1988, and is submitted pursuant to Section 7.5.1.e of the Fort St.
Vrain Technical Specifications.
Please contact Mr. M. H. Holmes at (303) 480-6960 if you have any questions regarding this report.
Very truly yours, R. O. Williams, Jr.
Senior Vice President, Nuclear Operations ROW:VHF/djc Attachment i
1 l
1 I
pf(
o
f l
f P-89073 February 28, 1989 cc w/ attachments:
Regional Administrator, Region IV U. S. Nuclear Regulatory Commission Attn: Mr. T. F. Westerman, Chief Projects Section B 611 Ryan Plaza Drive, Suite 1000 Arlington, TX 76011 Mr. Robert Farrell Senior Resident Inspector Fort St. Vrain Station Building 104 Dottie Sherman, ANI Library American Nuclear Insurers The Exchange, Suite 245, 270 Farmington Avenue Farmington, CT 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 Fort St. Vrain NED-Site Mr. Milt Lammering, Chief Regional Representative, Radiation Program U. S. Environmental Protection Agency Region VIII 999 18th Street, Suite 500 Denver, CO 80202-2405 i
Mr. Farrel D. Hobbs Environmental Management i
Rockwell International Building T-452-B P. O. Box 464 Golden, CO 80401
P-89073 February 28, 1989 Dr. James E. Johnson Radiology & Radiation. Biology. Dept.
135 BRB Colorado State University Fort Collins, CO 80523 i
n t.-
_ _ _ _ _ _ - - - _. _ _ _

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