ML20101F186

Annual Radioactive Effluent Release Rept for Jan-Dec 1995

ML20101F186
Person / Time
Site:	Fort Saint Vrain
Issue date:	12/31/1995
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML20101F178	List: ML20101F175 ML20101F186 ML20101F209
References
NUDOCS 9603250376
Download: ML20101F186 (19)

Text

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ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT January - December 1995 4

3 Public Service Company of Colorado i

Fort St. Vrain Nuclear Station March, 1996 1

9603250376 960321 PDR ADOCK 05000267 R PDR

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SUMMARY

This report summarizes radiological effluent released from the Fort St. Vrain Nuclear Station for the period of January through December, 1995 consistent with 10 CFR 50.36a. This information is provided pursuant to the requirements of Section 5.5.2 of the Fort St. Vrain Decommissioning Technical Specifications (DTS).

This report uses the reporting format recommended by Regulatory Guide 1.21 where appropriate.

The following tables with a supplemental information section are included with this report:

Table Descriotion 1A Gaseous Effluents - Summation of All Releases 2A Liquid Effluents - Summation of All Releases 2B Liquid Effluents - Batch and Continuous 3 Solid Waste and Irradiated Fuel Shipments Please note that Tables 1B and 1C (cf Regulatory Guide 1.21) have been omitted from this report because all of our gaseous effluents are continuous and assumed to be ground-level releases as opposed to being elevated releases.

Table 4A has also been omitted. The Offsite Dose Calculation Manual (ODCM), copy attached, has been revised to delete the requirement to collect meteorological data. Meteorological data is not needed for calculating dose contributions from effluent releases and it is not needed by the NRC's contractor who evaluates data submitted with the Annual Radiological Effluent Release Reports.

The ODCM relies upon a conservative default value of X/Q for dose calculations instead of using actual meteorological data. The annual average X/Q value of 4.59E-04 sec/m 3 is specified in the ODCM and in the Fort St. Vrain Decommissioning Plan. This value is conservative when compared to actual meteorological data (1989 actual X/Q = 1.95E-04). Also, the use of an annual average X/Q for release calculations is consistent with Reg Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gasaous Effluents in Routine Releases from Light-Water-Cooled Reactors,"

section C.4.

The Fort St. Vrain Decommissioning Technical Specificatiols and ODCM requirements apply to all reactor produced radionuclides that 2

are identifiable. Individual radionuclides are analyzed and included in this report.

Sample activities that are less than the detection limits are listed as "none detected" and no value is entered into the calculations of the total activities. This results in the reporting of values which are representative of the total curies released.

The lower limit of detection (LLD), for the purposes of this report, is defined as the lowest concentration of radioactive material in a sanple 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 the ODCM are as follows:

Liould Principal Gamma Emitters 5.00E-07 yCi/r.1 Tritium 1.00E-05 pCi/ml Gaseous Principal Gamma Emitters 1.00E-11 pCi/cc (Particulate)

Tritium (Gas) 1.00E-06 pCi/cc All samples were analyzed to the LLD concentration limits for nuclides specifically listed in our ODCM, however, those which had values of less than LLD were marked as "none detected". The "none detected" values were not included in the total values for the pathway.

With respect 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 miles downstream of the plant, no further dilutions were assumed. Additional discussion on river flow is contained in Section 4d of the Supplemental Information Section.

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

Table 3 summarizes the radioactive solid waste shipments for burial made during 1995. Since the table states " ....for burial or disposal" we have identified all shipments made to a burial site.

The summary includes shipments for burial performed by processors 3

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t and agents from their facilities, in addition to those made

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directly from Fort St. Vrain during 1995. Materials which were shipped to processors and agents for volume reduction, incineration, etc. in 1995 are not specifically identified in this '

report. The waste resulting from these processes, that was shipped for burial in 1995, is included in Table 3.

There were no major design changes to the radioactive liquid waste system reviewed by the Decommissioning Safety Review Committee (DSRC) during the period covered by this report. However, the  ;

holdup and clean up sections of the radioactive liquid waste system l were dismantled and decommissioned during this reporting period. '

The liquid waste sump, associated pumps, the liquid effluent '

activity monitors, and the existing liquid effluent piping remain in service to process and discharge the remaining liquid waste that will be generated as a result of decommissioning activities.

The rubber membrane liner holding tank installed in 1994 and described in the 1994 annual report (P-95023) also remains in service.

Issue 7 of the Decommissioning ODCM (DPP-5.4.2 attached to this  ;

report), was reviewed by the Decommissioning Safety Review Committee (DSRC) on October. 11, 1995. The revisions maintained the '

required levels of radioactive effluent control and did not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

There were no changes made to the Process Control Program during this reporting period.

l There were no liquid effluent activity monitors or their associated  !

recorders inoperable for more than thirty days during this reporting period. i l

Radiation doses resulting from the release of radioactive liquid i and gaseous effluents from Fort St. Vrain during this period are ,

reported below. Radiation doses were calculated in accordance with i the Fort St. Vrain ODCM which is based on NUREG-0472, " Radiological i' Effluent Technical Specifications for PWRs", NUREG-0133,

" Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", and other inputs from the NRC.

l Doses were calculated for a " maximum exposed" individual present at i all times of the year at the Emergency Planning Zone (EPZ), which j is a minimum of 100 meters from the Reactor Building exhaust stack, and located in the sector in which all of the activity was calculated to have been released.

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i The following are the 1995 doses:

Licuid - 10CFR50 Whole Body 1.40E-03 mrem Maximum Exposed Organ 7.07E-03 mrem Gaseous - 10CFR50 Noble Gas Gamma 0.00 mrem Noble Gas Beta 0.00 mrem Iodine, Particulates, Tritium Adult Whole Body 2.07E-01 mrem Organ (maximum) 2.07E-01 mrem Bone 0.00 mrem Teen Whole Body 2.37E-01 mrem Organ (maximum) 2.37E-01 mrem Bone 0.00 mrem Child Whole Body 3.40E-01 mrem Organ (maximum) 3.40E-01 mrem Bone 0.00 mrem Infant Whole Body 2.65E-01 mrem Organ (maximum) 2.65E-01 mrem Bone 0.00 mrem Gaseous - 10CFR20 Iodine, Particulates, Tritium 2.67E-01 mrem All doses are within the limits of 10CFR20 and 10CFR50.

The doses demonstrate conformance with the exposure limit in 40CFR190 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 released from Fort St. Vrain was 0.00 Ci. The 29.78 kev iodine-129 peak is below the minimum detectable energy of our detectors. However, waste streams are regularly analyzed specifically for iodine-129 for compliance with 10CFR61. The results of these analyses have been less than the LLD.

An annual land use census was performed as part of the Fort St.

Vrain Radiological Environmental Monitoring Program. Changes made to environmental sampling locations as a result of the land use census are reported in the Annual Radiological Environmental Monitoring Program Report (Annual Radiological Environmental Operating Report).

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Effluent and Waste Disposal Annual Report Sunclemental Information Facility: Fort St. Vrain Nuclear Generatina Station Licensee: Public Service Comoany of Colorado l

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1. Reaulatory 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 (ODCM) (DPP-5.4.2) to assure that the concentrations at the point of release are maintained within the regulatory limits contained in the ODCM. These limits 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. l l

The dose to a member of the public due to I-131, tritium, and l radioactive particulates with half-lives longer than eight i days in gaseous effluents will be limited to: i a) 7.5 millirems to any organ during any calendar quarter, and, b) 15 millirems to any organ during any calendar year.

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

a) During any calendar quarter to less than or equal to 1.5 millirems 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.

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2. Maximum Permissible Concentrations (MPC's)

' All MPC's used:in determining allowable release rates from the gas and liquid waste systems were those listed in Table II, Columns 1 and 2 respectively, of Appendix B to 10CFR20.

Public Service Company continues to use Table II of 10CFR20 because we were granted an exemption from the revised 10CFR20 requirements for the duration of the Fort St. Vrain decommissioning project.

3. Averace Enerav The average energy (E-Bar) of the radionuclide mixture in releases of fission and activation gases was not calculated or '

used at this facility.

4. Measurements and ADoroximations of Total Radioactivity a) Fission and Activation Gases Gaseouc releases from Fort St. Vrain consist of water  :

vapor that is evaporated from the surface of the shield water in.the PCRV, where it is drawn into the Airborne  :

Contamination Control System (ACCS), and then is  ;

exhausted via the Reactor Building ventilation exhaust '

system. This is a continuous release consisting of tritiated water vapor. This release methodology is acceptable because of the low activity level of the vapor being released from the PCRV.

Batch releases of radioactive gaseous waste were discontinued after March, 1993, when the gaseous waste hold up system was taken out of service.

All radioactive gases were released to the Reactor Building exhaust ventilation system which has a flow rate that can be varied from approximately 10,000 to 35,000 cubic feet per minute. . The full-flow of this exhaust was directed through high efficiency particulate filters l (HEPA) prior to the release to the environment.

1 Downstream of the HEPA filters, the gas stream I

radioactivity was sampled and analyzed on a weekly basis, b) Iodines The reactor was shut down permanently in August,1989 and the fuel removed completely in June of 1992 resulting in the elimination of the iodine source term. Therefore, iodine-131 analysis has been discontinued in accordance-with the ODCM.

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c) Particulates A 2-inch particulate filter was removed and analyzed each week. Gross beta analysis, as well as gamma spectral analysis, was performed to identify and quantify any radionuclides. The quantity of any radionuclides on this filter with half-lives greater than eight days was similarly correlated to total flow during the collection period.

d) Licuid Effluents All liquid effluent discharged from the site reached the i unrestricted area at the Goosequill Ditch. From that  ;

point, the effluent can be diverted to the St. Vrain i Creek via the St. Vrain Slough or, more commonly,  !

diverted to the Farm Pond which is approximately one mile i north. of the plant site. Outfall from the Farm Pond '

reaches the South Platte River. Both rivers converge approximately 1-1/2 miles from the plant site. The average stream flow reported in section 5a of this supplemental report is a summation of both rivers, and  ;

was tabulated from data provided_ by the Colorado Department of Natural Resources in Greeley, Colorado.

Liquid effluent was released from the site using a batch mode from the Reactor Building Sump gnd the radioactive liquid waste system. Liquid waste f- t the rubber liner ,

membrane tank was processed thrbugh f1'1ters and demineralizers, then directed to the Reactor Building sump for eventual discharge through the existing effluent pathway. ,

A sample of the tanks or sump contents was taken and analyzed prior to each release. These analyses include beta, tritium, and gamma spectroscopy. The results of j these analyses, and other analyses as dictated by the  :

gross beta results, were used to determine the maximum release rate from the site. The liquid effluent was diluted with cooling tower blowdown, which runs at or more than 1100 gallons per minute, prior to reaching the unrestricted area. The resulting mixture was sampled during the release period to confirm compliance with regulatory limits.

The liquid effluent from the batch release mode was monitored continuously by a gamma activity monitor.

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

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5. Batch Releases a) Liould Number of Batch Releases 27 Total Time Period for Batch Releases 9.11 E+02 HOURS Maximum Time Period for a Batch Release 1.66 E+02 HOURS Average Time Period for  !

a Batch Release 3.37 E+01 HOURS Minimum Time Period for a Batch Release 1.80 E+00 HOURS Average Stream Flow During Periods of Release of  ;

Effluent into a Flowing 4 Stream 7.07E+05 GPM*

• Gallons Per Minute b) Gaseous

1. None I

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6. Abnormal Releagga a) Licuid Number of Releases Total Activity Released Ci b) Gaseous Number of Releases Total Activity Released Ci 11

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I j' i TABLE 1A 4

EFFLUENT AND WASTE DISPOSAL j 4

GASEOUS EFFLUENT - SUMMATION OF ALL RELEASES l t

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Estimated )

I Unit Quarter Quarter Total i 1 2 Error l A. Fission and activation gases

i. 1. None detected i

l B.- Iodine ,

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1. None Detected C. Particulates i

1. None Detected D. Tritium

1. Total Release Ci 5.45E-01 9.33E-01 1.73E+01 l

2. Average release rate gCi/sec 7.01E-02 1.19E-01 for period

3. Percent of  % <1 <1 applicable limit 1

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TABLE 1A EFFLUENT AND WASTE DISPOSAL i GASEOUS EFFLUENT - SUMMATION OF ALL RELEASES i

Estimated  ;

Unit Quarter Quarter Total

. 3 4 Error -

A. Fission and activation gases

1. None detected.

B. Iodine i

1. None detected.

C. Particulates

1. None Detected D. Tritium i

1. Total Release Ci 1.32 E+00 9.47E-01 1.73E+01 l, i i

2. Average release rate gCi/sec 5.60E-02 4.01E-02  ;

i for period

3. Percent of  % <1 <1 applicable limit 4

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TABLE 2A EFFLUENT AND WASTE DISPOSAL LIQUID EFFLUENT - SUMMATION OF ALL RELEASES 4

Est.

Units Qtr 1 Qtr 2 Total l Error A. Fission and Activation Products 4

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, 1. Total Release Ci 1.24E-03 6.72E-02 1.73E+01

2. Average diluted 4Ci/ml 1.65E-08 8.52E-08 concentration

3.  % of Applicable  % <1 <1 Limit B. Tritium i

1. Total Release Ci 3.32E-01 1.74E+00 1.73E+01

2. Average diluted 4Ci/ml 1.26E-05 4.41E-06 concentration

3.  % of Applicable  % <1 <1 Limit C. Gross Alpha Radioactivity

1. None Detected D. Volume of Waste Released prior to dilution Liters 1.64E+05 6.61E+05 1.00E+01 E. Volume of Dilution Water Used During Release Liters 2.33E+07 3.26E+08 1.00E+01 14

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TABLE 2A EFFLUENT AND WASTE DISPOSAL LIQUID EFFLUENT - SUMMATION OF ALL RELEASES Est..  !

Units Qtr 3 Qtr 4 Total '

Error A. Fission and Activation Products

1. Total Release Ci 2.54E-01 8.69E-03 1.73E+01l

2. Average diluted gCi/ml 3.66E-07 5.51E-07 concentration

3.  % of Applicable  % <1 <1 Limit B. Tritium ,

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1. Total Release Ci 1.51E+00 4.01E-02 1.73E+01l 1

2. Average diluted gCi/ml 1.65E-05 2.82E-06 l concentration

3.  % of Applicable  % <1 <1 Limit C. Gross Alpha Radioactivity

a. None Detected D. Volume of Waste Released prior to dilution Liters 4.73E+05 1.37E+04 1.00E+01l E. Volume of Dilution Water Used During Release l Liters 8.10E+07 1.39E+07 1.00E+01l j i

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TABLE 2B '

EFFLUENT AND WASTE DISPOSAL l LIQUID EFFLUENTS l l

CONTINUOUS MODE BATCH MODE j Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 cobalt-60 Ci *

• 3.04E-06 2.15E-05 l

chlorine-36 Ci 1.23E-03 6.71E-02 cesium-137 Ci 3.09E-06 1.04E-04 i  !

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tritium Ci 3.32E-01 1.74E+00 l 1

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, Total for period Ci 3.33E-01 1.81E+00 1

(above) 4 1 o continuous liquid effluent discharges were not made during this reporting period.

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TABLE 2B EFFLUENT AND WASTE DISPOSAL LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE Nuclides Unit Quarter 3 Quarter 4 Quarter 3 Quarter 4 Released chlorine-36 Ci *

• 5.20E-02 1.49E-03 cobalt-60 Ci ND** 1.04E-06 iron-55 Ci 2.02E-01 7.20E-03 tritium Ci 1.51E-00 4.01E-02 l 1

1 i Total for period Ci 1.76E+00 4.88E-02 (above)

o Continuous liquid effluent discharges were not made during this reporting i period.

00 None Detected I

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4 TABLE 3

EFFLUENT AND WASTE DISPOSAL (1995) l SOLID WASTE AND IRRADIATED FUEL SHIPMENTS A. SOLID WASTE SHIPPED OFFSITE FOR BURIAL OR DISPOSAL (not irradiated fuel)

1. Type of waste Unit Period Est. Tot.

Error, %

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a. Spent resins, filter sludges, m3 3.13 E+01 6.44 E+00

. evaporator bottoms, etc. Ci 7.46 E-01 5.05 E+00 i b. Dry compressible waste, contaminated m3 4.42 E+02 3.00 E+00 equip, etc. ci 2.51 E+01 5.01 E-01

, c. Irradiated components, control rods, m8 1.32 E+03 3.00 E+00

] etc. Ci 4.12 E+02 1.46 E+00

d. Other (describe): Waste processed by a 1.03 E+02 1.00 E+01 vendor C1 2.92 E+01 1.00 E+01 i

i j 2. Estimate of major nuclide composition (by type of waste) in Curies 9

. a. Fe 7.16% 5.34 E-02

co-60 3.50% 2.61 E-02

! cs-134 6.78% 5.06 E-02

] Cs-137 77.48% 5.78 E-01 H-3 3.82% 2.85 E-02

b. Fe-55 73.71% 1.85 E+01 co-60 10.60% 2.66 E+00 Ni-63 1.38% 3.47 E-01 Eu-152 2.81% 7.06 E-01 H-3 10.16% 2.55 E+00

c. Fe-55 36.85% 1.52 E+02 co-60 12.29% 5.07 E+01 Ni-63 2.62% 1.08 E+01 H-3 43.88% 1.81 E+02 Eu-152 3.52% 1.45 E+01

d. Fe-55 21.22% 5.07 E+00 Co-60 4.35% 1.04 E+00 H-3 42.27% 1.01 E+01 __

Cs-137 1.24% 2.96 E-01 Eu-152 29.55% 7.06 E+00 Eu-154 1.82% 4.34 E-01 18

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3. Solid Waste Disposition Number of Shiomants Mode of Transoortation Destination 151 Public Highway US Ecology, Inc.  ;

Richland, WA l J

B. IRRADIATED FUEL SHIPMENTS (Disposition)

Number of Shioments Mode of Transoortation Destination None I

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