ML20081J783
| ML20081J783 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/31/1994 |
| From: | PUBLIC SERVICE CO. OF COLORADO |
| To: | |
| Shared Package | |
| ML20081J780 | List: |
| References | |
| NUDOCS 9503280179 | |
| Download: ML20081J783 (20) | |
Text
-
n P
I e
I ANNUAL' RADIOACTIVE EFFLUENT RELEASE REPORT January - December 1994 t
Public Service Company of Colorado Fort St. Vrain Nuclear Station March, 1995 r
1 4
9503280179 941231 PDR ADOCK 05000267 PDR R
1.O
SUMMARY
Th'is report summarizes radiological effluent released from the Fort St.
Vrain Nuclear Station for the period of January through December, 1994 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:
1 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 (of Regulatory Guide 1.21) have i
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 8
annual average X/Q value of 4.59E-04 sec/m 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 total 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 Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," section C.4.
The Fort St. Vrain Decommissioning Technical Specifications and ODCM requirements apply to all reactor produced radionuclides that 2
1 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 this 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 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 the ODCM are as follows:
Liauid Principal Gamma Emitters 5.00E-07 pCi/ml 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 dcwnstream of the plant, no further dilutions were assumed.
Additional discussion on river ~ 'ficW is contained in Section 4d of the Supplemental Information Fection.
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 made during 1994.
Since the table states "....for burial or disposal" we have identified only those shipments made to a burial site.
The summary includes shipments performed by processors and agents in 3
4 addition to those made directly by the station during 1994.
Materials which were shipped to a processor for volume reduction in 1994 are not included in this report.
There was'1 major design change to the radioactive liquid waste i
system reviewed by the Decommissioning. Safety Review Committee (DSRC) during the period covered by this report.
In accordance with decommissioning procedure DPP 4.1.1, Design / Design Document i
Control, changes to components / systems addressed by the DTS or by the ODCM are considered upgraded design and hence must be authorized by'an Engineering Change Request (ECR).
1.
To assist with lowering the Prestressed Concrete Reactor Vessel (PCRV) shield water level, a rubber membrane liner holding tank was installed inside the " key-way" below the shield water pumps and filters located on level 1 (Elevation 4740') in the Reactor Building.
The rubber liner tank was sized to contain about 35,000 gallons of water, drained from the PCRV, that could be stored for additional treatment prior to making a direct discharge into the Reactor Building Sump and a subsequent liquid effluent release.
In addition, the open topped rubber liner tank was designed so that the water could be recirculated through a newly installed line back into the shield water demineralizers.
This new recirculation pathway is for further water clean up if required prior to release.
The rubber liner and the new recirculation pathway were installed under ECR-94-004 and by means of a Maintenance Work Request (MWR).
Issue 5 of the Decommissioning ODCM (DPP-5.4.2 attached to this report),
was reviewed by the Decommissioning Safety Review Committee (DSRC) on September 15, 1994 and became effective September 19, 1994.
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.
There were no liquid effluent activity monitors or their associated recorders inoperable for more than thirty days during this reporting period.
Radiation doses resulting from the release of radioactive liquid and gaseous effluents from Fort St. Vrain during this period are reported below. Radiation doses were calculated in accordance with the Fort St. Vrain ODCM which is based on NUREG-0472, " Radiological Effluent Technical Specifications for PWRs",
" Preparation of Radiological Effluent Technical Specifications for i
Nuclear Power Plants", and other inputs from the NRC.
4 i
2 y,
I Doses were calculated' for a " maximum exposed"' individual present at
- all times of the year at the Emergency Planning Zone (EPZ),.which is a minimum of 100_ meters from the Reactor Building exhaust stack j
and located in the sector in which all of the activity was calculated to have been. released.
The following are the.1994
. doses:
Liauid - 10CFR50 t
Whole Body 5.13E-04 aren
.j Maximum Exposed Organ 3.29E-03 aren e
Gaseous - 10CFR50 Noble Gas Gamma-0.00 aren Noble Gas Beta 0.00 arem Iodine, Particulates, Tritium Adult Whole Body 1.04E-01 aren Organ (maximum) 1.04E-01 aren Bone 0.00 mrea i
Teen Whole Body 1.19E-01 aren Organ (maximum) 1.19E-01 aren Bone 0.00 aren Child Whole Body 1.71E-01 mram Organ (maximum) 1.71E-01 aren Bone 0.00 mren-Infant Whole Body 1.33E-01 arem Organ (maximum) 1.33E-01 aren Bone 0.00 mram Gaseous - 10CFR20 Iodine, Particulates, Tritium 1.34E-01 aren 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 arem 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. The reactor has been shutdown since August, 1989.
Any iodine-131 which was produced by the fission process has undergone in excess of 200 5
half-lives decay.
It is assumed the release of iodine-131 to the environment would be impossible, therefore, iodine-131 analysis has been discontinued in accordance with the ODCM.
An annual land use census was perforned 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).
6
Effluent and Waste Disnosal Annual 1Renort
^
.Sunnlemental Information Facility:
Fort St. Vrain Nuclear Generatina Station Licensee:
Public Service Connany of Colorado 1.
Reculatory 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.
The dose to a member of the public due to I-131, tritium, and radioactive particulates with ' half-lives longer than eight days in gaseous effluents will be limited to:
a) 7.5 millirems. to any organ. during any calendar quarter, and, b) 15 millirems to any organ dtiring any calendar year.
i 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 millirens 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.
i
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7
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I 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
)
Gaseous releases from Fort St. Vrain consist of water vapor that is evaporated from the surface of the shield j
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 (HEPA) prior to the release to the. environment.
Downstream of the HEPA
- filters, the gas stream 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 May of 1992 resulting in the elimination of the iodine source term.
Therefore, iodine-131 analysis has been discontinued in accordance with the ODCM.
8
I{
y c)
Particulateg l
i 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 i
radionuclides. The quantity of.any radionuclides on this filter with half-lives' greater than eight days was
.l similarly correlated to total-flow during the collection.
i period.
I d)
Liauid Effluents All liquid effluent discharged from the sitie reached the l
unrestricted area at the Goosequill Ditch.
From-that l
point, the affluent can be diverted to the St.:Vrain Creek via the St.
Vrain Slough or, more commonly, diverted to the Fara Pond which is approximately one mile 1
north of the' plant site.
Outfall from the Fara 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 Sa' of - this.
supplemental report is a summation of both rivers,;and was received and tabulated from data provided by the Colorado Department of Natural Resources in Greeley, Colorado.
Liquid effluent was released from the site using a batch' mode from the Reactor Building Sump and the radioactive-liquid waste system.
The Reactor Building Sump area can-hold several hundred thousand gallons of waste water from various' sources which could be contaminated.
The liquidi waste system consists of two'2000. gallon receivers, one 2000 gallon monitoring tank, and associated filters and' demineralizers.
This system was designed to collect and process contaminated waste water resulting from reactor =
and laboratory operations.
A sample of the tank or sump contents was taken and analyzed prior to each release.
These analyses include gross beta, gross alpha, tritium, and gamma spectroscopy.
The results of these analyses,>and other analyses as dictated by the gross beta results, were used to j
determine the maximum release rate from the site.
The H
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.
9
-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 i
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, 7
sampling
- errors, and tank volume estimates, each considered to be'plus or minus 10 percent.
i 1
'5:
10
)
.O.
- 5. -
Batch Releases i
1 a)
Liquid Number of Batch Releases 93 Total Time Period for Batch Releases 413.73 HOURS a
Maximum Time Period for a Batch Release 45.30~
HOURS Average Time Period for a Batch Release 4.45 HOURS Minimum Time Period for a Batch Release 0.37 HOURS i
Average Stream Flow During i
Periods of Release of Effluent into a Flowing Stream 2.61E+06 GPM*
- Gallons Per Minute l
b)
Gaseous i
1.
None i
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i I
f 1
b b
11
.h s-e
.e e
7
.;..V V :
=6.
Abnormal Releases a)
Liquid Number of Releases Total Activity Released Ci -
b)
Gaseous
-l'l Number of Releases Total Activity Released Ci a 1
?
i I
?
? \\r I
b i
i I
i l
12
i l
t TABLE 1A EFFLUENT ~AND WASTE DISPOSAL GASEOUS EFFLUENT - SUMMATION OF ALL RELEASES
-)-
Estimated Unit Quarter
-Quarter Total 1
2 Error iJ A.
Fission and activation _ gases 1.
None detected B.
Iodine i
1.
None Detected i
C.
Particulates i
t 1.
None Detected D.
Tritium l
l
- 1. Total Release Ci 1.93E-01 1.40E-01 1.73E+01
- 2. Average release rate pCi/sec 2.45E-02 1.79E-02 i
for period 3.
Percent of Technical
<1
<1 Specification Limit f
f 13 t
i TABLE 1A EFFLUENT AND WASTE DISPOSAL GASEOUS EFFLUENT - SUMMATION OF ALL RELEASES I
i I
Estimated Unit Quarter Quarter Total i
3 4
Error A.
Fission and activation gases 1.
None detected.
B.
Iodine i
1.
None detected.
C.
Particulates i
1.
None Detected i
D.
)
l'
- 1. Total Release Ci 4.21E-01 1.05E+00 1.73E+01
- 2. Average release rate pCi/sec 6.43E-02 1.32E-01 for period 3.
Percent of Technical
<1
<1 Specification Limit 14
e TABLE'2A I
EFFLUENT.AND WASTE DISPOSAL LIQUID EFFLUENT - SUMMATION OF ALL RELEASES l
Est.
l Units Qtr 1 Qtr Total Error 4
'A.
Fission and' Activation Products 1.73E+01l l
- 1. Total Release Ci 3.06E-04 1.04E-04 6
- 2. Average' diluted pCi/ml 9.09E-09 6.63E-09 concentration 3.
% of Applicable
<1 -
1 Limit l
B.
Tritium t
1.73E+01l
- 1. Total Release Ci 3.29E-01 3.52E-02
- 2. Average diluted pCi/ml 2.62E-05 2.23E-06'-
concentration 3.
% of Applicable
<1
<1 Limit C.
Gross Alpha Radioactivity I
1.
None Detected D.
Volume of Waste Released l prior to dilution Liters 2.39E+05 1.88E+05 1.00E+01l E.
Volume of Dilution Water Used During Release l
Liters 2.62E+07 5.32E+06 1.00E+01l 15
TABLE 2A EFFLUENT AND WASTE DISPOSAL LIQUID EFFLUENT - SUMMATION OF ALL RELEASES i
Est.
Units Qtr 3 Qtr 4 Total Error I
?
A.
Fission and Activation Products 1.73E+01l
- 1. Total Release Ci 2.78E-03 8.29E-03
- 2. Average diluted pCi/ml 2.61E-08 4.00E-08 concentration 3.
% of Applicable
<1
<1 Limit B.
Tritium 1.73E+01l
- 1. Total Release C1 1.26E-01 5.91E-01
- 2. Average diluted pCi/ml 3.68E-06 9.74E-06 concentration 3.
% of Applicable
<1
<1 Limit C.
Gross Alpha Radioactivity a.
None Detected D.
Volume of Waste Released 1.00E+01l prior to dilution Liters 3.48E+05 3.41E+05 E.
Volume of Dilution Water Used During Release 1.00E+01l Liters 5.24E+06 4.04E+06 16 l
TABLE 2B EFFLUENT AND WASTE DISPOSAL LIQUID EFFLUENTS CONTINUOUS MODE BATCH MODE Nuclides Released Unit Quarter 1 Quarter 2 Quarter 1 Quarter 2 cobalt-60 Ci 1.08E-04 4.74E-05 l
cesium-134 Ci 1.26E-05 ND**
cesium-137 Ci 1.70E-04 3.61E-05
?
europium-152 Ci ND ND europium-154 Ci 1.47E-05 1.92E-05 l!
l tritium Ci 3.29E-01 3.52E-02 Total for period Ci 3.29E-01 3.53E-02 (above) i Continuous liquid effluent discharges were not made during this reporting period.
None Detected i
i 17
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 ND**
4.57E-03 cobalt-60 Ci 1.90E-03 2.76E-03 cesium-134 Ci 2.24E-05 2.0AJ-05 cesium-137 Ci 3.90E-04 4.81E-04 europium-152 C1 2.68E-04 2.84E-04 europium-154 C1 1.99E-04 1.79E-04 l
l tritium Ci 1.26E-01 5.91E-01 l
Total for period Ci 1.29E-01 5.99E-01 (above) o Continuous liquid effluent discharges were not made during this reporting
)
period.
00 None Detected j
l I
l l
l 18 1
l i
.. ~. _
1 3
.l 1
TABLE 3 EFFLUENT'AND WASTE DISPOSAL (1994)
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.
Total l
Error, t
=
3
- a. Spent resins, filter sludges, m
3.04 E+01 evaporator bottoms, etc.
Ci 9.77 E-01 1.00 E+01 3
- b. Dry compressible waste, contaminated m
1.26 E+02 equip, etc.
CL 1.43 E+01 1.00 E+01 3
- c. Irradiated components, control rods, m
2.38 E+02 etc.
C1 3.84 E+03 1.00 E+01 3
- d. Other (describe):
m 0.00 E+00 Ci O.00 E+00 0.00 E+00 Irradiated concrete
- 2. Estimate of major nuclide composition (by type of waste) in Curies A. Fe-55 10.794 1.05 E-01 l
Co-60 3.80%
3.72 E-02 Cs-134 6.46%
6.32 E-02 l
Cs-137 74.12%
7.24 E-01 H-3 3.64%
3.55 E-02 t
- b. Fe-55 35.62%
5.10 E+00
-l Co-60 5.80%
8.31 E-01 t
Cs-137 9.104 1.30 I+00 Eu-152 17.964 2.57 E+00 Eu-154 1.104 1.57 E-01 H-3 29.60%'
4.24 E+00 l
t
7.31 E+02 Ni-63 2.20%
8.46 E+01 H-3 29.154 1.12 E+03 i
Eu-152 5.54%
2.13 E+02 Eu-154 3.80s 1.46 E+02
- d. N/A 19
4 :.
r 3.; Solid Waste Disposition-M =har of Shinments MAxie of Transoortation. Dgstination d
1 69' Public Highway US Ecology, Inc.
Richland, WA iB. IRRADIATED. FUEL SHIPMENTS (Disposition)
Number of Shinnents Mode of Trensoorta (gg gagg.ination None 5
b t
4 i
4 T
4 r
6 I
- l
' l, 20 l
e.
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