ML031270563
| ML031270563 | |
| Person / Time | |
|---|---|
| Site: | Cooper  (DPR-046) |
| Issue date: | 05/01/2003 |
| From: | Coyle M Nebraska Public Power District (NPPD) |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NLS2003051 | |
| Download: ML031270563 (76) | |
Text
-
I INFORMATION ONLY Uncontrolled Copy Assessment Manual
-ODAM-For assessment of Gaseous and Liquid Effluents at COOPER NUCLEAR STATION Brownville, Nebraska 28872 wPD
COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL
-ODAM-LIST OF EFFECTIVE PAGES Page Rev/Date Page Rev/Date Paqe Rev/Date ii 8126/99 8/26199 1
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(Deleted) 1/27/89 Appendix B:
C-5 C-6 C-7 C-8 C-9 Appendix D:
iii D1.0-1 D2.0-1 D3.0-1 D3.0-2 D3.1-1 D3.1-2 D3.1-3 D3.1-4 D3.1-5 D3.1-6 D3.1-7 D3.1-8 D3.1-9 D3.1-10 D3.2-1 D3.2-2 D3.2-3 D3.2-4 D3.2-5 D3.2-6 D3.2-7 D3.2-8 D3.2-9 D3.2-1 0 D3.2-11 10/10/01 0
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2/06/02 2/06/02 2/06/02 8/26/99 8/26/99 B-1 B-2 B-3 B4 B-5 B-6 B-7 Original Original Original Original Original Original Original Appendix C:
C-1 C-2 C-3 C-4 Original 2/06/02 2/06/02 2/06102 NOTE: Original refers to page in effect on July 1, 1986 when the ODAM was implemented.
Page 1 of 2 03/20103
COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL
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Page 2 of 2 03/20103
COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL FOR GASEOUS AND LIQUID EFFLUENTS Page 1.0 Introduction..........................................................
. 2.0 Liquid Effluent.................................................. 2.1 Radioactivity In Liquid Waste 2-2.2 Aqueous Concentration 2-2.3 Method of Establishing Alarm Setpoints.........
4-2.3.1 Setpoint for a Batch Release..............................-.
5-2.3.2 Setpoint for a Continuous Release..........................-.
8-2.4 Radioactivity Concentration in Water Beyond the Site and Exclusion Area Boundary.
10-2.5 Accumulated Personal Maximum Dose...........................-.
12-2.6 Proiected Personal Maximum Dose..............................-.
15-3.0 GaEseous Effluent.....................................................
3.1 Introduction.
3.2 Radioactivity in Gaseous Effluent.
3.3 Main Condenser Air Eiector Noble Gas Monitor Alarm Setpoint.
3.4 Effluent Noble Gas Monitor Alarm Setpoint.
3.4.1 Setpoint Based on Dose Rate.............................
3.5 Noble Gas Gamma Radiation Dose Accumulated in Air.
3.6 Noble Gas Beta Radiation Dose Accumulated in Air.
3.7 Dose Due to Iodine and Particulates in Gaseous Effluents*.
3.7.1 GASPAR Method......................................
3.7.2 Alternate Method......................................
3.8 Dose to a Person from Noble Gases. 26/99 8/
-i-
INFORMATION ONLY 3.8.1 Gamma Dose to Total Body - GASPAR Method........-.......
34-3.8.1.1 Alternate Method..........
.................-. 35-3.8.2 Dose to Skin - GASPAR Method
.................-........... 36-3.8.2.1 Alternate Method..........
.................-. 36-3.9 Proiected Organ Dose Due to Gaseous Effluent.........
...-......... 37-3.10 Dose Rate Due to Tritium. lodines. and Particulates in Gaseous Effluents 4.0 Dose Commitment From Releases Over Extended Time.
40-4.1 Releases During A Quarter 40-4.2 Releases During 12 Months 42-4.2.1 Calculated Doses.
43-4.2.2 Environmental Measurements.
45-5.0 Radiological Environmental Monitoring Program.
47-5.1 Environmental Sampling Program.
47-APPENDIX A - (Deleted)
APPENDIX B - Reference Meteorological Data APPENDIX C - Environmental Radiation Monitoring Program Sample Types and Sample Station Locations APPENDIX D - ODAM Specifications 8/26/99
J Wi kIt Ad tfgi X> Z S to OFFSITE DOSE ASSESSMENT MANUAL FOR GASEOUS AND LIQUID EFFLUENT 1.0 Introduction This Manual describes acceptable methods of calculating radioactivity concentrations in the environment and the potentially resultant personal dose equivalent commitment beyond the site and exclusion area boundary that are associated with LWR liquid and gaseous effluents.
The radioactivity concentrations and dose estimates are used to demonstrate compliance with the Appendix D Specifications required by 10 CFR 50.36.a. The methodology stated in this Manual is acceptable for use in demonstrating operational compliance with 10 CFR 20.1302, 10 CFR 50 Appendix I, and 40 CFR 190.10(a). Only the dose attributable to the Station is considered in demonstrating compliance with 40 CFR 190 since no other nuclear facility exists within 50 miles of the Station.
Calculations are made to assess the air dose from radioactive noble gases near ground level beyond the site and exclusion area boundary location that could be occupied by a person where the maximum air dose is expected. The maximum dose commitment to the person beyond the site and exclusion area boundary potentially experiencing the maximum exposure to all other radioactive material measured in gaseous and liquid effluents released from the Station is also calculated. Alternatively, the dose commitment from effluents other than radioactive noble gases may be calculated to correspond with residence at an occupiable location where airborne exposures are unlikely to underestimate those experienced by the maximally exposed person.
-I-10/15/96
I
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, i'-'
i, I o;,
II 4 _ U" '6 i 14,
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Y I
2.0 Liquid Effluent 2.1 Radioactivity In Liquid Waste The concentration of radionuclides in liquid waste is determined by sampling and analysis in accord with Table D3.1.1-1, Radioactive Liquid Waste Sampling and Analysis. Alternatively, pre-release analysis of the radioactivity concentration in liquid waste required by DSR 3.1.1.1 may be done by gross P-y counting provided an effluent concentration beyond the site and exclusion area boundary for unidentified emitters, 1 x 10.8 jiCilml, is applied where the discharge canal meets the river. When a radionuclide concentration is below the LLD for the analysis, it is not reported as being present in the sample. A general diagram of the liquid effluent stream components is shown in Figure 3.1A.
2.2 Aqueous Concentration Radioactive material in liquid effluent is diluted successively by water flowing in the discharge canal and in the river. The diluted concentration of radionuclide i in a receiving stream is estimated with the equation Fc C. = C where C.
C.
F, F2
=
concentration of radionuclide i
in liquid radwaste released (,iCi/ml) concentration of radionuclide i in the receiving stream (ItCi/ml)
=
release rate of liquid radwaste (ml/sec)*
=
dilution flow of receiving stream of water (mlIsec)*
- F,, F2, and F, may have any convenient units of flow (i.e., volume/time) provided the units of all are identical. 8/26/99
For the purpose of calculating the radioactivity concentration in water beyond the site and exclusion area boundary (Section 2.4), the flow in the discharge canal, F., is assigned to F2.
This method of estimating concentration of radionuclide i in a receiving stream is very conservative as it is based on not exceeding the concentration limits in IOCFR20 Appendix B, Table 2, Column 2 during the period of radioactive material discharge.
As an alternate to the above method, the concentration of radionuclide i in a receiving stream can be calculated using the monthly and quarterly composite samples. This method, discussed in the basis of DSR 3.1.1.2, is performed as follows:
For Sr-89, Sr-90 and Fe-55 Czi = CqciVrq Vdq where Czi Cqci Vrq Vdq
=
concentration of radionuclide i in the receiving stream (11ci/ml) concentration of radionuclide i in the quarterly composite sample (jici/ml)
=
volume of liquid radwaste discharged during the quarter (ml)
=
total volume of dilution flow corresponding to the time when Vrq was discharged for the quarter (ml)
I A
$ [t I 1-r 4.:q..p P2
_. -r-0
.1, 1 F V A '*j ; tI 1 I A-$ j j Q" .4 0 N, LY For all other nuclides Czi = Cmrci Vrm Vdm where Czi Cmci Vrm Vdm
=
concentration of radionuclide i in the receiving stream (ptci/ml)
=
concentration of radionuclide i in the monthly composite sample (Itci/ml)
=
volume of liquid radwaste discharged during the month (ml)
=
total volume of dilution flow corresponding to the time when Vrm was discharged for the month (ml) 2.3 Method of Establishing Alarm Setpoints The liquid waste effluent monitor and the service water monitor are connected to alarms which provide automatic indication when 10 CFR Part 20, Appendix B, Table 2, Column 2 concentrations are expected to be exceeded beyond the site and exclusion area boundary. With prompt action to reduce radioactive releases following an alarm, the liquid release limit of 10 CFR Part 20.1302 and the limits provided by 10 CFR Part 50, Appendix I,Section IV are unlikely to be exceeded after the alarm.
The alarm setpoint for the liquid effluent radiation monitor is derived from the concentration limit provided in 10 CFR Part 20, Appendix B, Table 2, Column 2 applied where the discharge canal flows into the river. The alarm setpoint does not consider dilution, dispersion, or decay of radioactive material in the river.
The radiation monitoring and isolation points are located in the liquid radwaste effluent line and the service water effluent line through which radioactive effluent is, or may be, eventually discharged into the discharge canal. 10/15/96
The alarm setpoint calculation for each liquid effluent monitor is based upon measurement according to Table D3.1.1-1 of radioactivity in a batch of liquid to be released or in the continuous aqueous discharge. Alternatively, the alarm setpoint may be based upon gross P-y activity analysis of the liquid waste provided the effluent concentration beyond the site and exclusion area boundary for unidentified emitters, 1 x 10.8 ItCi/ml, is observed.
In any case, a monitor may be set to alarm or trip at a lower activity concentration than the calculated setpoint.
2.3.1 Setpoint for a Batch Release A sample of each batch of liquid radwaste is analyzed for I-131 and principal gamma emitters, or for total activity concentration prior to release. The ratio, FMPCb, of the activity concentration in the tank to the effluent concentration (10 CFR Part 20, Appendix B, Table 2, Column 2) beyond the site and exclusion area boundary is calculated with the equation FMPC E
Cbpi identified bp MPC. 10/15/96
where FMPCbP
- Cbp, MPC,
=
fraction of effluent concentration beyond the site and exclusion area boundary in batch derived from activity measured prior to release.
=
concentration of radionuclide i (including I-131 and principal gamma emitters) in batch sample taken prior to release (iCi/ml).
=
effluent concentration beyond the site and exclusion area boundary of radionuclide i per 10 CFR Part 20, Appendix B, Table 2, Column 2 (RCi/ml)
When FMPCbp is derived from analyses identifying iodine and principal gamma emitters only, the value FMPCbP may be adjusted to account for radionuclides measured in the monthly and quarterly composite sample, but not measured prior to release. This adjustment, derived from measurements during past calendar quarters, is calculated with the equation:
FMPCb = FMPCbP - Eb where Eb =
(Previous quarterly average of the fraction of the effluent concentration in the discharge canal due to I-131 and primary gamma emitters) -
(Previous quarterly average of the fraction of the effluent concentration in the discharge canal due to all radionuclides in batch releases.)
A reference value of Eb, derived from representative past measurements may be used routinely.
Whether radioiodine and primary gamma emitters are identified prior to a batch release or not, the liquid radwaste effluent line radiation monitor alarm and isolation valve closure setpoint is determined with the equation:
S=
AP S-*
g + Bkg FMPCb F.,! 8/26/99
where S
=
radiation monitor alarm setpoint (cpm or [tCi/ml)
A counting rate (cpm/ml) or activity concentration ([tCi/ml) of sample from laboratory analysis*
g
=
ratio of effluent radiation monitor counting rate to laboratory counting rate or activity concentration in a given batch of liquid (cpm per jtCi/ml or iLCi/ml per ItCi/ml)
Fs1
=
maximum flow in the batch release line (gallmin)**
FS2
=
minimum flow in the discharge canal (gallmin)**
Bkg
=
monitoring instrument background (cpm or [iCi/ml)
Note that A-FMPCb represents the counting rate of a solution having the same radionuclide distribution as the sample and having the maximum permissible concentration of that mixture.
Gross P-y analysis alone may be used to determine the radioactivity in a batch prior to release. In that event, the fraction of the effluent concentration beyond the site and exclusion area boundary in the batch is:
FMPCbP Cbp-1 X lo-where CbP=
gross or total radioactivity concentration in batch sample taken prior to release (pCi/ml) 1 x 108
=
effluent concentration beyond the site and exclusion area boundary of unidentified radionuclides (ItCi/ml)
- A equals A, C.,, if isotopic analysis was performed or C bp if gross activity analysis was performed.
- Any suitable but identical units of flow (volume/time). 8/26/99
The value of FMPCbp computed with this expression is substituted in the preceding equation to calculate the setpoint.
2.3.2 Setpoint for a Continuous Release Continuous aqueous radioactive discharges are sampled and analyzed according to the schedule in Table D3.1.1-1. The ratio FMPCCW, of the activity concentration in each of the continuous release streams of the effluent concentration beyond the site and exclusion area boundary is calculated with the equations.
FMPC
= -
cwI identified cw 7
MPC.
where
- FMPC,
=
fraction of effluent concentration beyond the site and exclusion area boundary in continuous release based upon activity measured in weekly composite sample(s).
Ci
=
concentration of radionuclide i (including I-131 and principal gamma emitters) in weekly composite sample(s)
(giCi/ml)
When FMPC, is derived from analyses of 1-131 and principal gamma emitters, it may be adjusted to account for radionuclides measured in the monthly and quarterly composite sample but not measured prior to release. Adjustment for radionuclides measured in monthly and quarterly composite samples but not in weekly composite samples is given by the equation FMPCC = FMPCCW Er 10/15/96
where E.=
(Quarterly average fraction of the effluent concentration in the discharge canal due to I-131 and primary gamma emitters measured in weekly composite sample of continuous releases during previous quarter) -
(Quarterly average fraction of the effluent concentration in the discharge canal due to all radionuclides in samples of continuous releases during previous quarter.)
A reference value of Ec, derived from representative past measurements, may be used routinely, instead.
The alarm setpoint of the radiation monitor on the discharge line is determined with the equation A
- 5. g + Bkg FMPCC FS1 where A
=
counting rate (cpm/ml) or activity concentration (piCi/ml) of weekly composite sample in the laboratory.
Terms g, Fs,, FS2, and Bkg are defined the same as in the setpoint equation for a batch release.
Gross P-y analysis alone may be used to determine the radioactivity in a liquid radioactive discharge. In that event, the fraction of the effluent concentration beyond the site and exclusion area boundary in a sample of the release is:
cI FMP CC,-
I X lo-,
-9 8/26/99
where C.,
gross or total radioactivity concentration in continuous aqueous release (pCi/ml) 1 x 10-8
=
effluent concentration beyond the site and exclusion area boundary of unidentified radionuclides (gCi/ml)
The value of FMPCC computed with this expression is substituted in the preceding equation to calculate the setpoint.
In the event a long-term trend is evident in setpoints derived from the weekly sample and a setpoint value can be derived from the aggregate of the weekly samples which appears to have less variability and to better represent the effluent, then the setpoint based on the combined, long-term data may be used.
2.4 Radioactivity Concentration in Water Beyond the Site and Exclusion Area Boundary DSR 3.1.1.2 requires that measured radioactivity concentrations in liquid releases be evaluated to verify that the activity concentration complied with Specification DLCO 3.1.1. Compliance with DLCO 3.1.1 is evaluated by calculating the average radioactivity concentration in water at the end of the discharge canal, expressed as a fraction of effluent concentration beyond the site and exclusion area boundary on the basis of measured release(s), per Table D3.1.1-1, of Fe-55, Sr-89, and Sr-90 averaged over no more than 92 days and other radionuclides averaged over no more than 31 days. 8126/99
~~'
1l.
The average concentration of radioactive noble gases in discharge canal water may be calculated separately as a fraction of the effluent concentration 2 x 10-4 uCi/ml, since the critical exposure pathway for it, immersion in water, differs from the critical exposure pathway for other radionuclides in water, which is via ingestion of the water.
The average concentration, expressed as a fraction of the effluent concentration, is calculated with the equation:
FMPC =
I_
k Ik 3785(TE-TB) k 2
- MPC, where FMPC =
fraction of the effluent concentration beyond the site and exclusion area boundary of a mixture of radionuclides in water (unitless, and should be limited to
- 1) 3785
=
conversion factor (ml/gal)
TE-TB
=
increment of time between beginning and ending period of interest during which the concentration is averaged (min)
F2k
=
flow of aqueous stream beyond the site and exclusion area boundary into which radioactive release represented by sample k is diluted, i.e., the discharge canal flow during the release represented by sample k (gal/min)
=
quantity of radionuclide i represented by sample k which is released as an effluent within the time boundaries TB and TE (pCi)
I 8/26/99
MPC1
=
maximum permissible concentration beyond the site and exclusion area boundary of radionuclide i per 10 CFR Part 20, Appendix B, Table 2, Column 2 (pCi/ml)
The data used to compute FMPC are measured by the radioactive liquid sampling and analysis program described in Table D3.1.1-1.
2.5 Accumulated Personal Maximum Dose DSR 3.1.3.1 requires the dose or dose commitment to a member of the public due to radioactive material released in liquid effluent to be calculated on a cumulative basis at least once every 31 days. The requirement is satisfied by computing the accumulated dose commitment to the most exposed organ and to the total body of a hypothetical person exposed by eating fish taken from the river beyond the site and exclusion area boundary near the discharge canal and drinking water taken from the river three miles downstream.
The accumulated dose commitment is computed at least once every 31 days, but may be computed as analyses becomes available. The dose will be calculated in accordance with Regulatory Guide 1.109, Revision 1, utilizing the LADTAP II computer code.*
The LADTAP II program is routinely used for calculating radiological dose assessments for inclusion in the CNS Radioactive Effluents Release Report.
- With quality factor for Tritium reduced from 1.7 to 1.0 per ICRP. 8/7/00
Site specific parameters input to LADTAP II are listed below. These parameters are included in the program calculations and are only changed as conditions and/or situations warrant.
o CNS effluent water flow in cfs, the average flow in the discharge canal during the time of interest.
o Dilution factor for the effluent.
Drinking water:
- 5 (for LADTAP variable Dilution factor)
Fish*: *5 (for LADTAP variable Dilution factor)
- Fishing - Seasonal variation: Consumption of fish is evaluated from April through November.
Alternatively, the accumulated dose commitment may be calculated in the following way:
ADk =3.785X10 e
eani Clk tk(
F D
AD an k
ank where AD.flk
=
the dose commitment (mrem) to organ n of age group a due to the isotopes in a release represented by analysis k, where the analyses are those required by Table D3.1.1-1.
Thus the contribution to the dose from gamma emitters become available on a batch basis for batch releases and on a weekly basis for 12129/88
continuous releases. Similarly the contributions from H-3 are available on a monthly basis and the contributions from Sr-89, Sr-90, and Fe-55 become available on a quarterly basis.
Dan
=
the dose commitment attributed to releases represented by all analyses k to organ n, including total body, of the maximally exposed person in age group a (mrem).
Aeai
=
transfer factor relating a unit release of radionuclide i (Ci) in a unit stream flow (gal/min) to dose commitment to organ n, or total body, of an exposed person in age group a via environmental pathway e mrem Ci m ini gal 3.785 x 103 = 3785 ml/gal x 106Ci/,ICi C
=
the concentration of radionuclide i in the undiluted liquid waste to be discharged (giCi/ml), i.e., in the sample k Atk elapsed time of release represented by sample k during which radionuclide i is discharged at concentration Cik, i.e., the duration of the release represented by sample k (minutes)
(F/IF2) =
the quotient of the release flow, F., and the dilution flow, F2, during the release represented by sample k 12/29/88
Pathway-to-dose transfer factors, A=ani for use in calculating the dose commitment arising from radioactive material released in aqueous effluents, are calculated in accordance with equations and values recommended in Regulatory Guide 1.109, Revision 0. Appropriate factors representing applicable environmental pathways of exposure and most exposed age group(s) are selected and used in calculating the dose commitment. The pathway(s) and thus age group(s) selected may vary by season. For instance, when fishing near the Station during the winter is nonexistent, evaluation of the fish pathway is not required.
The age group potentially most exposed via eating fish is expected to be the adult, and the age group potentially most exposed via drinking water from the Missouri River is expected to be the infant. Normally, only these need to be evaluated for compliance with DSR 3.1.3.1. For the purpose of calculating the dose to the Member of the Public who is potentially exposed most by eating fish taken from the river beyond the site and exclusion area boundary near the discharge canal, F2 = 5Ff. As long as potable water is known not to be taken from the river within three miles downstream of Cooper Station, as verified by the annual land survey, the potential dose to a Member of the Public via drinking water will be assessed on the basis of water assumed to be taken from the river three miles downstream. At that location, F2 is conservatively assumed to be F2 = 5Fr. Variables F,, F2, and F, are defined in Section 2.2.
2.6 Projected Personal Maximum Dose DSR 3.1.3.2 requires the maximum total body and organ doses to a person beyond the site and exclusion area boundary due to radioactive material released in liquid effluent to be projected over a quarter at least one time during every 31 days if
_15_
05/16/94
RI-Vi A'~
radioactive liquid radwaste is released and the radwaste system is not operated.
This requirement is satisfied by calculating the projected dose commitment to a hypothetical person exposed by eating fish taken from the river beyond the site and exclusion area boundary near the discharge canal and drinking water taken from the river three miles downstream. The potential dose commitments to organs and to the total body are computed separately.
The quarterly dose commitment to a maximally exposed hypothetical person is projected by computing the accumulated doses to the total body and most exposed organ during the most recent three months and assuming the result represents the projected doses during the current quarter. Doses will be calculated in accordance with Section 2.5.
As an alternative, the quarterly dose commitment to the total body and most exposed organ may be projected by using the equation P"=91 L Pan = XDan where Pa.
=
projected dose commitment (mrem) to organ n (including total body of age group a during the current quarter) 91
=
number of days in a quarter X
=
number of days to date in current quarter Dan
=
dose commitment during the quarter-to-date (mrem) based upon results of aqueous effluent sampling and analyses available to date during the quarter 8/26/99
3.0 Gaseous Effluent 3.1 Introduction The Station discharges gaseous effluent through a stack (Elevated Release Point) and discharges ventilation air from the radwaste, augmented radwaste, turbine, and reactor buildings through the respective building vents.
These gaseous effluent streams, radioactivity monitoring points, and effluent discharge points are shown schematically in Figure 3-1. Gaseous release point locations and elevations at Cooper Station are described in Table 3-1. Gaseous discharges from the Elevated Release Point (ERP) are treated as an elevated release while discharges via building vents are assumed to be ground-level releases or split-wake releases.
Gaseous release point locations and elevations at the Station are described in Table 3-1.
3.2 Radioactivity in Gaseous Effluent For the purpose of estimating radionuclide concentrations and radiation doses, beyond the site and exclusion area boundary measured radionuclide concentrations in gaseous effluent and in ventilation air exhausted from the Station are relied upon.
Table D3.2.3-1 identifies the radioactive gaseous effluent measurements.
When a radionuclide concentration is below the LLD for the analysis, it is not reported as being present in the sample.
i ih A.3iEi".'
-t '
\\
+
Noble Gases. The distribution of noble gas radionuclides in a gaseous effluent is determined in one of the following ways.
- 1.
Preferably, the radionuclide distribution is obtained by gamma spectrum analysis of effluent gas samples in accordance with DLCO 3.2.3, Table D3.2.3-l. Results of analyses of one or more samples may be averaged to obtain a representative spectrum.
- 2.
In the event a representative radioactive noble gas distribution is unobtainable from samples taken during the period of interest, it may be derived from previous measurements or may be based upon a computed spectrum appearing in Table 3-2.
- 3.
Alternatively, the total activity concentration of radioactive noble gases may be assumed to be krypton - 88.
The total quantity of radioactive noble gas discharged during an interval of time is determined by integrating the rate measurement of each effluent noble gas monitor.
This may be done by the effluent monitoring system or the measured activity discharged via a gaseous effluent stream may be calculated with the equation Q=2.8 x 10 4 F
g where Q
=
total radioactive noble gas release via a gaseous effluent stream during a given time interval (jICi) t N
=
net counts accumulated during the time interval g
=
effluent noble gas monitor counting rate response
( pCi/cm3)
F
=
gaseous effluent stream discharge rate (cfm) 2.8 x 104 = conversion constant (cm3/ft3 )
3.3 Main Condenser Air Ejector Noble Gas Monitor Alarm Setpoint A noble gas activity monitor is provided to measure gross gamma activity in gases at the main condenser air ejector. The monitor includes an alarm that is set to report when the gamma radiation level in gas discharged by the main condenser air ejector indicates the gross radioactivity discharge rate exceeds 1 Ci/sec.
The alarm setpoint is determined with the relation S = 2120
+ Bkg F
where S
=
the alarm setpoint (mrfhr) h
=
monitor response to activity concentration of SJAE offgas being monitored (mr/hr per (pCifcm3 ))
F
=
air ejector discharge rate (cfm)
Bkg
=
monitoring instrument background (mr/hr)
P
=
fraction of allowable limit representing a chosen margin of conservatism in the setpoint (unitless) 2120 = 1 Ci x 1. OE6pCi x 1 ft3 x
60 sec Sec.
Ci 28317 cm3 min.
3.4 Effluent Noble Gas Monitor Alarm Setpoint DSR 3.3.2.9 requires an alarm setpoint to be determined for each radioactive noble gas effluent monitor. Each setpoint is derived to cause the alarm to report when the dose equivalent rate beyond the site and exclusion area boundary due to radioactive noble gas in gaseous effluent exceeds a limit in DLCO 3.2.1.a. Each noble gas activity monitor included in Table D3.3.2-1 except the main condenser air ejector off gas monitor is set to initiate alarm at or below the derived setpoint.
For the purpose of deriving a setpoint, the distribution of noble gas radionuclides in an effluent stream is determined as described in Section 3.2.
3.4.1 Setpoint Based on Dose Rate The alarm setpoint of a radioactive noble gas effluent monitor may be calculated on the basis of whole body dose equivalent rate beyond the site and exclusion area boundary. A setpoint of a monitor of an elevated release, e.g., from the stack, may be calculated with the equation. 10/15/96
E c, S = 1.06 h-P l
+ Bkg 7
F(C
- DFB) i The setpoint of a monitor of a ground-level or split-wake release, e.g., from the turbine building vent or the AOG building, may be calculated with the equation E C, h-P i
S 1.06
+ Bkg Q
DFv) i where S
=
the alarm setpoint (cpm, mr/hr, or [iCi/cm3) h
=
monitor response to activity concentration of effluent being monitored, (cpm per giCi/cm3, mr/hr per [tCi/cm3, or ptCi/cm3 per ILCi/cm3)
C,
=
relative concentration of noble gas radionuclide i in effluent at the point of monitoring (jiCi/cm3)
X/Q
=
atmospheric dispersion from point of ground-level or split-wake release to the location of potential exposure (sec/m3)
DFj
=
factor converting elevated release rate of radionuclide i to total body dose equivalent rate at the location of potential exposure (mrem)/(yr -yCi/sec) 8/26/99
DFj
=
factor converting ground-level or split-wake release of radionuclide i to the total body dose equivalent rate at the location of potential exposure (mrem)/(yr pCi/m3) f
=
flow of gaseous effluent stream, i.e., flow past the monitor (ft3/min)
Bkg
=
monitoring instrument background (cpm, mr/hr, or jiCi/cm3) 1.06
=
500 mrem/yr x 60 sec/min x 35.3 ft3/m3 x 1.0 m3/1.0 x 106 cm3 P
=
fraction of allowable limit representing a chosen margin of conservatism in the setpoint (unitless)
Each monitoring channel has a unique response, h, which is determined by the instrument calibration. In order to ensure the correct derivation of a setpoint, the monitor background (Bkg) and the monitor response factor (h) must be in consistent units.
The concentration of each noble gas radionuclide i in a gaseous effluent is determined as discussed in Section 3.2. 8/26/99
The atmospheric dispersion and the dose conversion factor, DFS depends upon local I
conditions. For the purpose of calculating radioactive noble gas effluent monitor alarm setpoints appropriate for Cooper Station, the locations of maximum potential exposure beyond the site and exclusion area boundary and the reference atmospheric dispersion factors applicable to the derivation of setpoints are:
Discharge Discharge Receptor Location Atm. Dispersion Point Height Sector Distance(m)
(sec/m3)
Vent Ground-Level NNW 1,150 3.4 x 10-6 or Split-Wake _
ERP Elevated W
1,800 8.2 x 10.8 The applicable dose conversion factors, DFSI and DFV, for deriving setpoints are in Table 3-3. 8/26/99
3.5 Noble Gas Gamma Radiation Dose Accumulated in Air DSR 3.2.2.1 requires the calculation on a cumulative basis of air dose due to gamma radiation from radioactive noble gas released in gaseous effluents. DLCO 3.2.2, Condition A requires reporting to the NRC when the air dose beyond the site and exclusion area boundary due to noble gas gamma radiation exceeds 5 mrad during any calendar quarter or 10 mrad during any calendar year.
The distribution of radioactive noble gases in gaseous releases and the quantity discharged during an interval of interest are determined as described in Section 3.2. 8/26/99 1
The gamma radiation dose to air beyond the site and exclusion area boundary as a consequence of noble gas released from the station will be calculated in accordance with Regulatory Guide 1 109, Revision 1, utilizing USNRC Computer Code GASPAR.*
The GASPAR program is routinely used for calculating radiological dose assessments for inclusion in the CNS Radioactive Effluents Release Report.
Site specific parameters input to GASPAR are listed below. These parameters are included in the program calculations and are only changed as conditions and/or situations warrant.
o Source terms for the intervals of interest o
Fraction of year milch animals are on pasture o
Atmospheric dispersion factors, for specific locations, obtained from annual meteorological data which are used to determine noble gas (gamma and beta) air doses for the following Maximum individual site boundary Maximum individual nearest resident Maximum individual nearest milch cow Maximum individual nearest garden o
Population distribution (0 to 50 miles) o Meat, milk, and vegetable distribution (0 to 50 miles) o Absolute humidity at Cooper Nuclear Station o
Fraction of the year vegetables are grown o
Fraction of daily feed intake derived from pasture for milch and meat animals
- Quality factor for Tritium reduced from 1.7 to 1.0 per ICRP. 8/7/00
Alternatively, the gamma radiation dose to air beyond the site and exclusion area boundary as a consequence of noble gas released from the station may be calculated with the equation:
D)9(QCS.AyCiJ
+Q i
(.J
. AV]
where D = noble gas gamma dose to air (mrad)
Q. = Z AQC5 = cumulative release of noble gas nuclide i from stack (jiCi).
1 1
Ay^
=
factor converting unit noble gas stack release to ground level air dose from overhead plume gamma radiation (mradlpCi).
Ay,
=
factor converting time integrated, ground level concentration of noble gas to air dose from gamma radiation 8126/99
=.3 AQcvi
=
cumulative release of noble gas nuclide i from building time vents (ILCi).
(X' long term average atmospheric dispersion factor for a ground level or
=Q) split wake release (sec/m3).
DSR 3.2.2-1 is satisfied by calculating the noble gas gamma radiation dose to air beyond the site and exclusion area boundary at a point situated at the NNW site boundary, and on the basis of reference* atmospheric dispersion assuming continuous gaseous release. At that location, the reference atmospheric dispersion factor for a vent (ground-level) release is X/Q = 3.4 x 10.6 sec/M3. Appropriate values of Ayc,, and Ay, for use in calculating air doses at that location are listed in Table 3-4.
3.6 Noble Gas Beta Radiation Dose Accumulated in Air DLCO 3.2.2 requires that the air dose beyond the site and exclusion area boundary from beta radiation not exceed 10 mrad during any quarter and 20 mrad during any year. DSR 3.2.2-1 requires the air dose to be calculated on a cumulative basis.
The radioactive noble gas distribution and activity discharged are determined as described in § 3.4 herein.
- Onsite meteorological data for the period July 1, 1976, to June 30, 1977, which was used in the Cooper Station Demonstrated of Compliance with 10 CFR 50, Appendix I, revision 1, January, 1978. 8/26/99
The beta radiation dose to air beyond the site and exclusion area boundary as a consequence of noble gas released from the station will be calculated in accordance with Regulatory Guide 1.109, Revision 1, utilizing USNRC Computer Code GASPAR.*
Alternatively, the beta radiation dose to air beyond the site and exclusion area boundary as a consequence of noble gas released from the station may be calculated with the equation E (Q X
+
X Ar1 Where D
=
noble gas beta dose to air (mrad)
(X/Q), =
long-term average atmospheric dispersion factor for stack releases (sec/m3)
Ap, factor converting time integrated ground level concentration of noble gas radionuclide i to air dose from beta radiation mrad
([tCi sec)lm3 DSR 3.2.2.1 is satisfied by calculating the noble gas beta radiation dose to air beyond the site and exclusion area boundary at a point situated at the NNW site boundary, and on the basis of reference atmospheric dispersion assuming continuous gaseous discharge. At that location, the reference atmospheric dispersion factors are:
(X/Q). =
1.2 x 10 8 sec/m3 at the NNW site boundary (X/Q)v =
3.4 x 106 sec/m3 Beta radiation-to-air dose conversion factors, AP,, for noble gas radionuclides are listed in Table 3-4.
- Quality factor for Tritium reduced from 1.7 to 1.0 per ICRP. 8/26/99
3.7 Dose Due to Iodine and Particulates in Gaseous Effluents*
DLCO 3.2.3 requires that radioiodine, and radioactive material in particulate form having half-lives greater than eight days in gaseous effluents released to the area beyond the site and exclusion area boundary cause no more than 7.5 mrem to any organ of a member of the public during any calendar quarter or 15 mrem to an organ of a member of the public during any calendar year. DSR 3.2.3.2 requires the dose to be calculated at least once every 31 days.
Radionuclides other than noble gases or tritium in gaseous effluents that are measured by the sampling and analysis program described in Table D3.2.3-1 are used as the release term in dose calculations. Airborne releases are discharged either via the stack (ERP) as an elevated release or via building vents and treated as a ground level or split-wake release. For each of these release combinations, samples are analyzed weekly, monthly, quarterly, or for a specific release according to Table D3.2.3-1.
Each sample provides a measure of the concentration of specific radionuclides, C, in gaseous effluent discharged at flow, F., during a time increment At. Thus, each release is quantified according to the relation:
- The dose to any organ of a person arising from radioactive iodine-13 1, iodine-133, and radioactive material in particulate form havinghalf-lives greater than eight days. Noble gases not considered.
8/26/99
At I
f
-, Ii
,4 AQjk -C&kFGJ Atj Q,k
] Ck FaJAtj where Qk CAk FAt Atj
=
the quantity of radionuclide i released in a given effluent stream based on analysis k (Ci)
=
concentration of radionuclide i in gaseous effluent identified by analysis k (Ci/mr) or (XCilcm3)
=
effluent stream discharge rate during time increment At)(m 3/sec)
=
elapsed time in increment j during which radionuclide i at concentration Cik is being discharged (see) 3.7.1 GASPAR Method A person may be exposed directly to an airborne concentration of radioactive material discharged in effluent and indirectly via pathways involving deposition of radioactive material onto the ground. Dose estimates account for the separate exposure pathways.
The dose commitment to a person beyond the site and exclusion area boundary associated with a gaseous release, Q,, of radioactive material other than noble gas will be calculated in accordance with Regulatory Guide 1.109, Revision 1, utilizing USNRC Computer Code GASPAR.
The GASPAR program is routinely used for calculating radiological dose assessments.
Site specific parameters input to GASPAR are listed in Section 3.5.
- Quality factor for Tritium reduced from 1.7 to 1.0 per ICRP. 8/26/99 1
3.7.2 Alternate Method Alternatively, the dose commitment to a person beyond the siteand exclusion area boundary associated with a gaseous release, Qk, of radioactive material other than noble gas may be calculated with one of the appropriate following equations release via the stack:
D4k i
( Q )c
(
Q eaeQ) bsej release via a vent:
D Xd~
A~i'
+ E TG (nt D
'nvk an T a
) b e
( Qle where Dansk
=
the dose commitment (mrem) to organ n of a person in age group a due to radionuclides identified in analysis k of an elevated (ERP) release where the analysis is one required by Table D3.2.3-1.
Danvk
=
the dose commitment from a vent release (mrem)
TAam
=
factor converting airborne concentration of radionuclide i to dose commitment to organ n of a person in age group a
( mrem '1 (Ci sec)fm3)
TGeau =
factor converting ground deposition of radionuclide i to dose commitment to organ n of a person in age group a exposed via environmental pathway e (mrem/Ci/m )
(D/Q) =
relative deposition factor (mn2) 8/26/99
(Xd/Q) =
depleted atmospheric dispersion factor (mCi/m3 per mCi/sec)
The analysis index k may represent either p, analysis of a grab sample w, a weekly composite analysis m, a monthly composite analysis q, a quarterly composite analysis The dose commitment accumulated by a person beyond the site and exclusion area boundary is computed at least every 31 days, but may be calculated as analytical results of effluent measurements, performed according to Table D3.2.3-1, become available.
The dose is accumulated in the following way.
The dose accumulated as a result of stack discharge is Dan Xv Dansw + m Dansm + q Daysq and the dose accumulated as a result of a vent discharge is Da.
W DIanvi m Danvm +q anvq Doses committed during the same time period due to discharges from the stack and vents are additive, thus:
Dan = Dans + v Da.,v where Dan
=
the dose commitment accumulated during the quarter to date as a result of all measured radioactive gaseous discharges except noble gases and tritium to any organ n, including total body, of a person offsite in age group a (mrem) 8/26/99
A 5 ;S >- _
When the dose to a person from iodine and particulates discharged in gaseous effluent is calculated as required by DSR 3.2.3.2, appropriate environmental pathways of exposure will be evaluated. The pathway(s) and/or age group(s) selected may vary by season. Appropriate pathway-to-dose transfer factors, Aea,,, are selected for use in calculating the dose.
The dose to a receptor 1.1 miles west of the Station is calculated on the basis of continuous gaseous release and reference meteorological conditions. The reference atmospheric dispersion and deposition factors at that location to be used for assessing compliance with DLCO 3.2.3 are:
(
= 8.1 x 10 sec/n 3 (2)
= 4.6 x 10-10 m 2 (XL = 4.4 x 10- sec 3
()
= 9.5 x io'l rn 2 The receptor is assumed to drink milk produced by the milch animal which experiences the maximum D/Q. Maximum values of the relative deposition factors for a milch animal located 3.7 miles northwest of the Station, are:
(D)
= 1.2 x 10-10 m-2 Q I (D)
= 3.7 x 10-10 Mr2 Qv 40 CFR Part 190. When the dose due to gaseous effluent is calculated for the purpose of evaluating compliance with 40 CFR Part 190 (reference Section 4.2), the dose contributed by tritium is included in the evaluation and is calculated in the following way. 8/26/99
Since tritium in water vapor is absorbed directly by vegetation, the tritium concentration in growing vegetation is proportional to the airborne concentration rather than to relative deposition as in the case of particulates. Thus the dose commitment from airborne tritium via vegetation (fruit and vegetables),
air-grass-cow-milk, or air-grass-cow-meat pathways is calculated with the appropriate one(s) of the equations:
for a stack release Dask =
i Qu TAanrp for a vent release Dani Q
=
I Q,,.
-TAanip 3.8 Dose to a Person from Noble Gases DSR 3.4.1.1 requires the calculation of dose to a member of the public for the purpose of assessing compliance with provisions of 40 CFR Part 190.10(a). That assessment includes the calculation of the gamma dose to the total body and the beta plus gamma dose to the skin of the person due to radioactive noble gases in gaseous effluents.
3.8.1 Gamma Dose to Total Body - GASPAR Method The gamma radiation dose to the whole body of a member of the public as a consequence of noble gas released from the station will be calculated in accordance with Regulatory Guide 1.109, Revision 1, utilizing USNRC Computer Code GASPAR.*
- Quality factor for Tritium reduced from 1.7 to 1.0 per ICRP. 8/26199 1
3.8.1.1 Alternate Method Alternatively, the gamma radiation dose to the whole body of a member of the public as a consequence of noble gas released from the Station may be calculated with the equation:
Dy
=
( lQCSPYCS +
(+
Y where Dy
=
noble gas gamma dose to total body (mrem)
Py;=
factor converting unit noble gas nuclide i in stack release to total body dose at ground level received from the overhead plume (mrem/fiCi)
Py,
=
factor converting time integrated, ground level concentration of noble gas nuclide i to air dose from gamma radiation
(.mrem jiCi sec m 3 When the total body dose due to gamma radiation from noble gas is evaluated as required by DSR 3.4.1.1, the dose to the nearby resident exposed most by all applicable exposure pathways combined is computed. Alternatively, the nearby resident exposed to maximal ground-level noble gas concentrations (maximum X/Q) may be selected as the receptor. Values by Pycs, and Pyvi for this receptor, 1.1 miles west of the station, appear in Table 3-5. 8/26199
3.8.2 Dose to Skin - GASPAR Method The beta radiation dose to the skin of a member of the public due to beta radiation from noble gas released from the station will be calculated in accordance with Regulatory Guide 1.109, Revision 1, utilizing USNRC Computer Code GASPAR.*
3.8.2.1 Alternate Method Alternatively, the beta radiation dose to the skin of a member of the public due to beta radiation from noble gas released from the Station may be calculated with the equation Dp=StQ X
Q X
- Spi DP (1.,
I
+ Q115 SR where DP
=
noble gas beta dose to skin (mrem)
S,
=
factor converting time integrated ground level concentration of noble gas radionuclide i to skin dose from beta radiation inrem (pCi sec)!M3 Values of Sp, for noble gases are included in Table 3-5.
When the skin dose due to noble gas beta radiation is evaluated as required by DSR 3.4.1.1, the receptor selected is the nearby resident exposed most via all applicable exposure pathways together. Alternatively, the nearby resident exposed to maximal ground-level concentrations (maximum X/Q) may be selected as the receptor. The location of the latter resident is 1.1 miles west of the station.
- Quality factor Tritium reduced from 1.7 to 1.0 per ICRP. 8/26/99
The total dose to the skin from noble gases is approximately equal to the beta radiation dose to the skin plus the gamma radiation dose to the total body.
3.9 Proiected Organ Dose Due to Gaseous Effluent DSR 3.2.5.1 and DSR 3.2.4.2 requires organ dose to a member of the public due to radioactive material in air effluent be projected during each month in which radioactive material is released in gaseous effluent without treatment. The purpose is to guide plant personnel in operating the EVTS and the Offgas Treatment System.
The organ dose is projected by calculating the dose to the most exposed organ accumulated during the month to date in accordance with Sections 3.7 and by projecting it for an entire 31 day time by employing the equation:
PD = 31 D where:
=
projected organ dose to a member of the public (mrem) 31
=
number of days over which dose is projected X
=
number of days to date during the projection period D
=
dose accumulated to the most exposed organ of a member of the public during the month to date (mrem). 8/26/99 I
3.10 Dose Rate Due to Tritium, lodines. and Particulates in Gaseous Effluents DLCO 3.2.1.b requires that the dose rate to any body organ created by the release of tritium, radioiodines, and radioactive material in particulate form having half-lives greater than eight days, shall not exceed 1500 mrem/yr. DSR 3.2.1.1 requires the dose rate to be calculated at least once every 31 days.
The dose equivalent rate from tritium, iodine, and radionuclides in particulate form in airborne effluent due to exposure by inhalation plus tritium absorption through the skin may be calculated for each discharge point by using the following equations. For effluent from an elevated release point, i.e., stack discharge above building wake, the equation is:
10-6 E E TE-TB k i
( )
TA For effluent from a ground-level release point, i.e., a building vent, the equation is:
D 10-6 E
(x'i any TE-TB k X
h ke Dose rates from separate release points may be combined to give Dan = Dan +
Danv where D.,,
=
dose equivalent commitment rate to organ n of a person in age group a due to radioactive particulates, iodine, and tritium in airborne effluent that are inhaled (mremfhr) 8/26/99
- 4.
,.L t
DanspDanv TE TB TE-TB (XTQ).e(XaQ),
TA,...,
=
dose equivalent commitment rate due to radioactive particulates, iodine, and tritium from an elevated release and a ground-level release respectively (mrem/hr)
=
quantity of radionuclide i released in a given effluent stream, either elevated or ground-level, based on analysis k (uCi) during discharge time increment TB to TE (hr)
=
ending time of effluent discharge
=
beginning time of effluent discharge
=
effluent averaging time (hr)
=
atmospheric dispersion from an elevated or a ground-level release respectively to ground-level at the receptor (uCi/m3 per uCi/sec)
=
factor converting airborne concentration of radionuclide i to dose commitment to organ n of a person in age group a and where e represents exposure via inhalation
(
nrem (Ci sec)Im3)
=
conversion, 10Ci/pCi 10.6 I
The analysis index, k, may represent either a grab sample, an integrated (continuous) sample, or a composite sample of an effluent. In turn, each sample represents certain radionuclides in the effluent during the time increment represented by the sample. 8/26/99
4.0 Dose Commitment From Releases Over Extended Time 4.1 Releases During A Quarter An annual assessment of radiation doses arising from liquid and gaseous effluents from the Station during each calendar quarter is required. The assessment includes the following calculations of doses for
- 1.
total body and maximally exposed organ doses due to liquid effluent via drinking water and eating fish from the river as in § 2.6.
- 2.
total body and maximally exposed organ doses due to gaseous effluents* other than noble gases and tritium as in § 3.7.
- 3.
doses to air offsite due to noble gas y as in § 3.5 and due to noble gas j3 as in
The dose calculations are based on liquid and gaseous effluents from the Station during each calendar quarter determined in accord with Tables D3.1.1-1 and D3.2.3-1.
- radioactive iodine-131, iodine-133, and radioactive material in particulate form, having half-lives greater than eight days. 8/26/99 1
3-
Aqueous concentration is estimated according to 2.2 on the basis of quarterly averaged stream flow or stream flow during discharge.
If practical, quarterly averaged meteorological conditions concurrent with the quarterly gaseous release being evaluated are used to estimate atmospheric dispersion and deposition. Otherwise, the quarterly dose commitment due to gaseous effluent will be calculated using either reference meteorology or annual averaged meteorology during the year in which the release occurred.
The receptor of the dose is described such that the dose to any resident near the Station is unlikely to be underestimated. That is, the receptor is selected on the basis of the combination of applicable pathways of exposure to gaseous effluent identified in the annual land use census and maximum ground level X/Q at the residence.
Conditions (i.e., location, X/Q, and/or pathways) more conservative (i.e. expected to yield higher calculated doses) than appropriate for the maximally exposed individual may be assumed in the dose assessment.
Seasonal appropriateness of exposure pathways may be considered. Exposure by eating fresh vegetation or drinking milk from cows or goats fed fresh forage is an inappropriate assumption during the first or fourth calendar quarter; rather consumption of stored vegetation and stored forage is ordinarily assumed.
Similarly, the liquid effluent-river-fish-man pathway is not ordinarilyassumed during the winter quarter.
8/26/99
Factors converting stack-released noble gas to gamma radiation dose from the overhead plume are calculated on the basis of reference meteorological data for the receptor location.
4.2 Releases During 12 Months The regulation governing the maximum allowable dose or dose commitment to a member of the public from all uranium fuel cycle sources of radiation and radioactive material in the environment is stated in 40 CFR Part 190.10(a). It requires that the dose or dose commitment to a member of the public from all sources not exceed 75 mrem/yr to the thyroid or 25 mrem/yr to the total body or any other organ.
DSR 3.4.1.1 requires calculation of the dose at least once per year to assess compliance with the regulation. If conditions warrant, according to provisions of DLCO 3.4.1, an assessment may be made for a portion of a calendar year.
Fuel cycle sources and nuclear power reactors other than the Station itself do not measurably or significantly increase the radioactivity concentration in the vicinity of the Station; therefore, only radiation and radioactivity in the environment attributable to the Station itself are considered in the assessment of compliance with 40 CFR Part 190.
8/26/99
/
The dose to a member of the public which is due to exposure to radioactive material in liquid and gaseous effluents from the station are ordinarily calculated while the dose attributable to irradiation is evaluated with environmental radiation dosimetry.
The receptor of the dose is selected on the basis of the combination of applicable pathways of exposure to gaseous effluent identified in the annual land use census and minimum atmospheric dispersion factor (maximum ground level X/Q) at his residence.
The receptor is described such that the dose to any resident near the Station is not likely to be underestimated. Conditions more conservative than appropriate for the maximally exposed (real) person may be assumed in the dose assessment.
4.2.1 Calculated Doses Doses to a member of the public are calculated on the basis of liquid and gaseous effluents from the station determined in accord with Tables D3.1.1-1 and D3.2.3-1.
Contributions to the dose due to liquid and gaseous effluent are calculated as described by the equation for:
- 1.
Totalbody and maximally exposed organ doses due to liquid effluent via drinking water and eating fish from the river as in § 2.6.
- 2.
Total body dose due to noble gas y as in § 3.8.1.
- 3.
Skin dose due to noble gas P as in § 3.8.2.
8/26/99
- 4.
total body and maximally exposed organ doses due to gaseous effluents* other than noble gases as in § 3.7.
Aqueous radioactive material concentrations are estimated according to § 2.2 on the basis of annual averaged stream flow.
Atmospheric dispersion, deposition, and if calculated, exposure by irradiation from airborne emitters are based on annual averaged meteorological conditions during the year evaluated or, alternatively, on reference meteorological conditions. In the event a portion of the year is examined, average meteorology for the period examined may be used in lieu of annual averaged or reference meteorology data.
Factors converting stack-released noble gas to gamma radiation dose from the overhead plume are calculated on the basis of annual averaged meteorological data for the receptor location.
- radioactive iodine, tritium, and radioactive material in particulate form having half-lives greater than eight days. 8/26/99 1
,A
4.2.2 Environmental Measurements When assessing compliance with 40 CFR 190, Radiological Environmental Monitoring Program results may be used to indicate actual radioactivity levels in the environment attributable to CNS as an alternate to calculating the concentrations from radioactive effluent measurements. The measured environmental activity levels may thus be used to supplement the evaluation of doses to real persons for assessing compliance with 40 CFR 190.
The dose to a member of the public due to irradiation (external exposure to gamma radiation) from the station and station effluents will be estimated with the aid of environmental TLD, PIC, or similar environmental dosimetry. This will be done by examining the annual dosimetry data for a statistical difference between measurements near the station and background measurements.
Alternatively, irradiation attributable to station effluents may be calculated by methods referenced earlier in this section.
The person most exposed to radiation and radioactive material in effluent from Cooper Station is expected to live within ten miles of the Station. Although the Station is in a rural area, the maximum personal exposure due to airborne effluent almost certainly occurs to a resident within three or four miles of it. Since the nearest public water intake downstream of Cooper Station in the Missouri River is about 85 miles, radioactive liquid effluent contamination of potable water is not foreseen to be significant. The other liquid effluent pathway of potential significance, via fish taken from the river, would be evaluated when assessing compliance with 40 CFR 190 only in the event that a significant increase in fishing downstream in the river near the Station occurs during the previous 12 months. Fishing within about ten miles
-45.
8/26/99
downstream of the Station is considered to be nonexistent during the first quarter and negligible during the remainder of the year. In the event the fish pathway is evaluated to assess compliance with 40 CFR 190, the fish would be taken from the river within ten miles downstream of the Station.
8126/99
I I
o 1
t s
f a+
5.0 Radiological Environmental Monitoring Program 5.1 Environmental Sampling Program DSR 4.1.1 requires a minimum radiological environmental monitoring program to be conducted as described in Table D4.1-1 of that document. APPENDIX C of the ODAM provides a numerical listing of the active sample stations along with a description of the sample types, locations, and maps showing their approximate location.
A radiological environmental monitoring program, approved by the Nuclear Regulatory Commission (NRC) was initiated at CNS before initial criticality was attained on February 21, 1974. The program monitors radiation levels in the air, terrestrial, and aquatic environments. Most samples are collected by Nebraska Public Power District (NPPD) personnel. However, all samples are shipped for analysis to a contractor's laboratory where there exists the special facilities required for measurements of extremely low levels of radioactivity. 8/26/99 Or
AD partulateauirlter H.
WMh elcW partiate air filter Atmosphere flaw Vent
[VTS F Vent 9'
Metir
..lHOEvtar 8 aJcbue MVTS
-0 Oab.cue Turbin SOuding
[
Hadweate BUMlin Potentially Contaminated Area Augmented Radwasto Bundd D
I 1) l IT
,I.
.,)
il FIgure 3-1 Gaseous EMuent Streams, Treatment and Monitoring Efoment and Dscharge Points.
OUl242,dlf 2/22/01 ;
TO DISCHARGE CANAL FIGURE 3.14 - LIQUID EFFLUENT STREAM, TREATMENT AND MONITORING,AND DISCHARGE POINT.
8/26/99
Table 3-1 Atmospheric Gaseous Release Points at the Cooper Nuclear Generating Station Combined Radwaste-Elevated Reactor Turbine Augmented Radwaste Release Structure Building Building Building Point Number of Ducts Duct Size (inches)
Height of Vent (feet above roof)
Flow Rate (cfm)
Flow Velocity (fps)
Exhaust - Winter Temp. (OF) - Summer Release Mode I
96" x 48" 15 73405 3.82 70 90 Partial Elevated 4
48" x 96" 1.3 5071 0(a) 26.4 70 90 Ground Level I
72" LI.D.
Horizontal discharge at rooftop 67070 39.5 70 90 Ground Level 1
14" I.D.
325 (above grade) 3000 46.7 60 90 Elevated (a)
Data given is for one operating fan. Multiply data by total number of fans in operation.
I.
I 1I, rr-em I 2/22/01
I Table 3-2 Computed Release of Radioactive Noble Gases In Gaseous Effluent From Cooper Nuclear Station Stack Release Plant Vents Release Nuclide (Ci/yr)
Fraction (Ci/yr)
Fraction Kr-83m 3.60E+01 8.38E-03 0
0 Kr-85m 6.50E+01 1.51E-02 7.10E+01 1.14E-02 Kr-85 2.OOE+02 4.66E-02 0
0 Kr-87 2.13E+02 4.96E-02 1.33E+02 2.13E-02 Kr-88 2.13E+02 4.96E-02 2.33E+02 3.74E-02 Kr-89 1.OOE+03 2.33E-01 0
0 Xe-133m 3.OOE 00 6.99E-04 0
0 Xe-133 1.51E+02 3.52E-02 2.63E+03 4.22E-01 Xe-135m 7.20E+01 1.68E-02 6.96E+02 1.12E-01 Xe-135 2.64E+02 6.15E-02 1.06E+03 1.70E-01 Xe-137 1.20E+03 2.79E-01 0
0 Xe-138 8.77E+02 2.04E-01 1.41E+03 2.26E-01 Total 4294.
1.0 6233.
1.0 Releases computed by BWR-GALE for Cooper Station Base Case gaseous radwaste treatment.
The release rate (Ci/yr) is included only to show the basis of the radionuclide distribution. To estimate the concentrations of radionuclides in a sample in which only the total radioactivity has been measured, multiply the total activity concentration by the fraction of respective radionuclides listed above.
8/26/99
Table 3-3 Dose Conversion Factors for Deriving Radioactive Noble Gas Effluent Monitor Setpoints Radionuclide Factor DFS, for Stack Release' (mrem)/(yr -Ci/sec)a mrem/gCia Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Kr-90 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Xe-139 Ar-41 3.5E-9 1.2E-4 1.7E-6
- 5. 1E-4 1.4E-3 6.6E-4 1.1E-16 3.8E-12 5.5E-14 1.6E-11 4.4E-11 2.1E-11 Factor DEj'for Ground-Level or Split-Wake Release (mrem)/(yr -pCi/m3) 7.56 E-2 1.17 E3 1.61 El 5.92 E3 1.47 E4 1.66 E4 1.56 E4 9.15 El 2.51 E2 2.94 E2 3.12 E3 1.81 E3 1.42 E3 8.83 E3 5.02 E3 8.84 E3 I
I 3.1E-5 2.3E-5 2.5E-5 2.5E-4 1.9E-4 5.4E-5 8.OE-4 1.6E-5 9.7E-4 9.7E-13 7.3E-13 8.OE-13 7.8E-12 6.OE-12 1.7E-12 2.5E-11 5.2E-13 3.1E-11 aBased on reference meteorology; applicable at the site boundary, ERP.
1,250 meters NNW of the 8/26/99
IL V
I t-Ln
,-:
It -
Table 3-4 Transfer Factors for Maximum Dose To A Person Beyond The Site And Exclusion Area Boundary Due To Radioactive Noble Gases Radionuclide Dose Transfer Factors a
- Aycs, Ayv, ApN mrad mrad mrad ItCi I C1s-ec/m 3 iCi sec/m3 I
Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Kr-90 Xe-13 lm Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Ar-41 2.6E-14 4.OE-12 5.8E-14 1.7E-11 4.6E-11 2.2E-11 1.1E-11 8.7E-13 9.OE-13 8.3E-12 6.3E-12 1.8E-12 2.7E-11 3.2E-11 6.1E-7 3.9E-5 5.4E-7 2.OE-4 4.8E-4 5.5E-4 5.2E-4 4.9E-6 1.OE-5 1.1E-5 1.1E-4 6.1E-5 4.8E-5 2.9E-4 2.9E-4 9.13E-6 6.24E-5 6.18E-5 3.26E-4 9.28E-5 3.36E-4 2.48E-4 3.52E-5 4.69E-5 3.33E-5 2.34E-5 7.79E-5 4.02E-4 1.51E-4 1.04E-4
'Dose at NNW site boundary 8/26/99
Table 3-5 Transfer Factors for Maximum Dose To A Person Beyond Site and Exclusion Area Boundary Due To Radioactive Noble Gases Radionuclide Dose Transfer Factors a,b Pycsi Pyvi Spi mrem mrem mrem pCji iCi sec/mr3 jCi sec/r 3 I
Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Kr-90 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Ar-41 1.6E-16 2.4E-12 3.OE-14 7.9E-12 2.3E-11 6.7E-12 7.7E-13 5.9E-13 6.9E-13 3.3E-12 3.7E-12 5.1E-13 1.2E-11 1.5E-11 2.4E-9 3.7E-5 5.1E-7 1.9E-4 4.7E-4 5.3E-4 4.9E-4 2.9E-6 8.OE-6 9.3E-6 9.9E-5 5.7E-5 4.5E-5 2.8E-4 2.8E-4 4.6E-5 4.2E-5
- 3. 1E-4 7.5E-5 3.2E-4 2.3E-4 1.5E-5 3.1E-5 9.7E-6 2.3E-5 5.9E-5 3.9E-4 1.3E-4 8.5E-5
'Receptor located at 1.1 miles west of Station bBased on reference meteorology at Cooper Station 8/26/99
APPENDIX A (DELETED)
A-1 1/27/89
APPENDIX B REFERENCE METEOROLOGICAL DATA Reference meteorological measurements were at Cooper Station during the period from July 1, 1976, through June 30, 1977. The summary data and the computer code, PUFF, were used to generate tables of reference values of X/Q, depleted X/Q, and DIQ herein.
B-i
UNDEPLETED MEAN RELATIVE CONCENTRATION (sec/m3)
ELEVATED RELEASE POINT-STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
SECTOR
.5 1.5 2.5 3.5 4.5 7.5
- 15.
- 25.
- 35.
45.
NNE 6.7E-09 2.3E-08 2.2E-08 1.8E-08 1.5E-08 1.9E-08 5.8E-09 4.7E.09 3.OE-09 1.8E-09 NE 6.1E-09 1.4E.08 1.4E.08 1.3E-08 1.1E-08 1.5E-08 6.9E-09 2.7E-09 2.4E-09 1.8E.09 ENE 7.OE-09 1.4E.08 1.4E-08 1.2E-08 9.3E-09 1.3E-08 2.9E-09 3.7E-09 1.5E-09 9.4E-10 E
6.5E-09 1.4E-08 1.3E-08 1.2E-08 9.5E-09 1.5E-08 4.OE-09 2.3E-09 1.3E-09 3.OE-10 ESE 5.2E.09 1.2E-08 1.OE-08 9.8E-09 7.9E-09 7.3E-09 4.lE-09 1.8E-09 1.2E-09 6.3E-10 SE 8.2E-09 1.9E-08 1.6E-08 1.4E-08 1.2E-08 l.OE-08 3.7E-09 1.6E-09 1.3E-09 6.5E-10 SSE 1.1E-08 3.2E-08 2.3E-08 2.OE-08 3.4E-08 2.6E-08 6.1E-09 2.2E-09 2.3E-09 1.2E.09 S
1.9E-08 3.4E-08 3.3E-08 2.6E-08 2.5E-08 1.6E-08 4.8E-09 2.4E-09 1.4E-09 1.1E-09 SSW 1.OE-08 4.3E-08 1.7E-08 1.7E-08 1.4E-08 9.5E-09 2.5E-09 1.2E-09 9.9E-10 5.1E.10 SW 4.4E-09 5.0E-08 1.7E-08 1.1E-08 1.1E-08 9.3E-09 3.1E-09 1.5E-09 9.4E-10 7.3E-10 WSW 4.1E-09 6.6E-08 3.2E.08 2.8E.08 1.2E-08 6.6E-09 4.1E.09 1.6E-09 1.1E-09 5.OE-10 W
5.6E-09 6.8E.08 3.8E-08 2.2E-08 1.8E.08 6.4E.09 4.1E-09 1.3E.09 8.2E-10 4.9E-10 WNW 6.1E-09 8.OE-08 5.2E-08 3.4E-08 2.1E-08 9.5E.09 3.2E.09 1.6E.09 1.OE-09 6.6E-10 NW 4.8E-09 8.8E.08 7.4E-08 5.2E.08 3.3E.08 1.4E-08 7.2E-09 3.4E-09 1.9E-09 1.3E-09 NNW 8.4E-09 2.7E-08 7.9E-08 6.9E.08 2.2E-08 2.1E-08 5.5E-09 3.1E.09 2.2E-09 1.6E-09 N
7.5E.09 3.5E-08 3.3E-08 2.5E-08 2.OE-08 1.6E-08 6.8E-09 5.2E.09 3.4E-09 1.1E-09
,4, IN B-2 I
UNDEPLETED MEAN RELATIVE CONCENTRATION (sec/m3 )
GROUND LEVEL RELEASE POINT - STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
SECTOR
.5 1.5 2.5 3.5 4.5 7.5
- 15.
- 25.
- 35.
45.
NNE 3.2E.06 5.5E-07 2.2E-07 1.5E-07 8.OE-08 4.4E.08 1.2E-08 4.9E-09 3.2E-09 2.4E-09 NE 2.OE-06 3.3E.07 1.8E.07 1.2E-07 6.1E-08 3.1E*08 9.2E-09 4.1E-09 2.6E-09 1.4E-09 ENE 2.2E-06 2.9E-07 1.5E-07 8.1E.08 5.4E-08 2.OE-08 7.4E-09 3.1E-09 1.6E-09 8.OE-10 E
2.2E-06 3.lE.07 1.5E.07 7.2E-08 5.5E-08 2.3E-08 6.3E-09 3.1E-09 1.8E-09 9.6E.10 ESE 2.4E-06 3.9E-07 1.5E-07 7.8E-08 5.7E-08 2.7E-08 7.4E-09 2.6E.09 1.3E-09 8.1E-10 SE 2.4E.06 3.9E.07 1.6E-07 1.2E.07 6.1EO08 2.5E-08 6.5E-09 1.8E-09 1.OE-09 7.8E-10 SSE 3.8E.06 6.OE-07 2.6E.07 1.5E-07 9.6E-08 4.2E-08 8.7E-09 2.8E-09 1.7E-09 1.2E-09 S
4.6E-06 8.lE-07 3.7E-07 2.OE-07 1.4E-07 6.6E.08 1.8E.08 6.4E-09 3.6E-09 2.1E-09 SSW 2.6E-06 5.OE-07 2.1E-07
- 1. lE-07 8.4E-08 5.5E-08 5.6E-09 1.5E.09 8.2E-10 4.8E-10 SW 1.9E-06 2.6E-07 1.8E-07 8.1E-08 6.2E-08 2.OE-08 5.2E-09 1.OE-09 3.9E-10 2.5E.10 WSW 2.OE-06 2.8E.07 1.7E-07 9.OE-08 6.4E-08 1.7E-08 3.6E-09 1.3E-09 7.4E-10 5.1E-10 W
1.6E-06 3.7E*07 1.4E.07 1.OE-07 6.5E.08 1.9E-08 6.1E-09 2.4E-09 1.1E-09 6.OE-10 WNW 3.1E-06 4.9E-07 2.2E.07 1.2E.07 1.OE.07 3.7E-08 1.OE-08 4.1E-09 2.1E-09 1.2E-09 NW 4.9E-06 7.8E.07 3.4E.07 2.2E.07 1.3E-07 6.5E-08 1.9E-08 5.OE-09 2.8E-09 2.OE.09 NNW 6.1E-06 9.7E.07 4.1E.07 2.5E.07 1.7E-07 9.5E-08 2.9E-08 1.2E.08 5.8E-09 1.6E-09 N
5.2E-06 8.9E-07 3:9E 07 2.2E-07 1.6E.07 7.4E-08 2.4EU08 1.lE-08 6.lE-09 3.5E-09
.4 I.
B.3
DEPLETED MEAN RELATIVE CONCENTRATION (sec/m3 )
ELEVATED RELEASE POINT-STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
SECTOR
.5 1.5 2.5 3.5 4.5 7.5
- 15.
- 25.
- 35.
45.
NNE 6.6E-09 2.2E.08 2.1E.08 1.7E.08 1.5E-08 1.8E-08 5.4E-09 4.5E-09 2.8E-09 1.6E-09 NE 6.OE-09 1.4E-08 1.4E-08 1.3E-08 1.1E-08 1.5E-08 6.5E.09 2.5E.09 2.2E.09 1.7E-09 ENE 6.9E-09 1.3E.08 1.4E-08 1.1E-08 8.8E-09 1.3E-08 2.7E-09 3.5E-09 1.4E-09 8.6E-10 E
6.4E-09 1.3E-08 1.3E-08 1.lE-08 9.0E-09 1.5E-08 3.9E-09 2.2E-09 1.2E-09 2.6E-10 ESE
- 5. 1E-09
- 1. 1E.08 1.OE.08 9.5E-09 7.6E-09 6.9E-09 3.9E-09 1.6E-09
- 1. 1E-08 9.6E-08 3.4E-09 1.4E-09
- 1. 1E-09 5.5E-10 SSE 1.1E.08 3.1E-08 2.3E-08 2.OE-08 3.3E-08 2.5E.08 5.6E-09 1.9E-09 2.OE-09 9.8E-10 S
1.9E-08 3.3E-08 3.2E-08 2.5E.08 2.4E-08 1.6E-08 4.4E-09 2.OE-09 1.1E-09 8.3E-10 SSW 1.OE-08 4.3E-08 1.7E.08 1.6E-08 1.4E-08 9.OE-09 2.3E-09 1.0E.09 8.6E-10 4.2E-10 SW 4.3E-09 4.9E-08 1.6E-08
- 1. 1E-08 1.OE-08 9.OE-09 2.9E-09 1.4E-09 8.4E-10 6.4E-10 WSW 4.OE-09 6.6E.08 3.2E.08 1.7E-08 1.1E-08 6.3E-09 3.9E-09 1.5E-09 9.5E-10 4.2E-10 W
5.5E.09 6.8E-08 3.7E-08 2.1E-08 1.7E-08 6.OE-09 3.8E-09 1.1E.09 6.8E-10 4.OE-10 WNW 6.OE-09 7.9E-08 5.1E.08 3.3E.08 2.1E-08 9.OE-09 3.OE-09 1.4E-09 8.8E-10 5.5E-10 NW 4.7E-09 8.7E-08 7.3E-08 5.1E-08 3.2E.08 1.3E-08 6.9E-09 3.1E-09 1.7E-09 1.2E-09 NNW 8.3E-09 2.6E.08 7.SE-08 6.8E-08 2.1E.08 2.1E.08 5.1E-09 2.8E.09 2.OE-09 1.5E-09 N
7.3E.09 3.5E.08 3.2E.08 2.4E.08 1.9E-08 1.5E-08 6.3E-09 4.8E-09 3.1E-09 9.4E-10
.1 It B-4 I
DEPLETED MEAN RELATIVE CONCENTRATION (sec/m3 )
GROUND LEVEL RELEASE POINT - STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
SECTOR
.5 1.5 2.5 3.5 4.5 7.5
- 15.
- 25.
- 35.
45.
NNE 2.8E.06 4.5E-07 1.7E-07 1.1E-07 6.1E-08 3.2E-08 7.8E.09 2.7E.09 1.6E-09 1.1E-09 NE 1.7E-06 2.8E-07 1.4E-07 9.1E-08 4.6E.08 2.2E-08 5.7E*09 2.2E-09 1.2E-09 5.6E-10 ENE 1.9E-06 2.4E-07 1.2E-07 6.2E-08 4.OE-08 1.4E-08 4.7E.09 1.7E-09 7.7E-10 3.3E-10 E
1.9E.06 2.5E-07 1.2E-07 5.5E-08 4.1E-08 1.6E-08 3.9E.09 1.5E-09 8.3E-10 3.9E-10 ESE 2.1E.06 3.2E-07 1.2E-07 6.OE-08 4.3E-08 1.9E-08 4.6E-09 1.5E-09 6.3E-10 3.9E-10 SE 2.1E-06 3.2E.07 1.3E-07 9.OE-08 4.6E-08 1.7E-08 3.9E-09 9.5E-10 5.OE-10 3.6E-10 SSE 3.3E*06 5.OE-07 2.1E.07 1.2E.07 7.3E-08 3.OE-08 5.4E-09 1.6E-09 8.5E-10 5.2E-10 S
4.OE-06 6.7E-07 3.OE-07 1.6E-07 1.lE-07 4.8E-08 1.2E-08 3.7E-09 1.9E 09 9.4E-10 SSW 2.3E-06 4.2E-07 1.7E 07 8.1E.08 6.3E-08 3.9E-09 3.4E-09 8.4E-10 4.2E-10 2.1E-10 SW 1.7E-06 2.2E.07 1.4E-07 6.1E-08 4.5E-08 1.4E.08 3.1E-09 5.8E-10 1.8E-10 1.1E-10 WSW 1.7E.06 2.3E-07 1.4E-07 6.8E.08 4.7E.08 1.2E.08 2.1E.09 7.OE-10 3.8E-10 2.5E-10 W
1.4E-06 3.OE-07 1.1E.07 7.7E-08 4.8E-08 1.3E-08 3.7E-09 1.2E-09 5.OE-10 2.7E-10 WNW 2.7E.06 4.OE-07 1.7E-07 9.2E-08 7.6E.08 2.7E-08 6.3E-09 2.3E.09 1.OE-09 5.8E-10 NW 4.1E.06 6.5E.07 2.7E-07 1.7E-07 1.OE-07 4.7E-08 1.2E-08 2.9E-09 1.5E-09 9.3E-10 NNW 5.4E-06 8.1E-07 3.3E-07 1.9E.07 1.3E-07 6.9E*08 1.9E-08 6.5E-09 3.OE-09 7.6E-10 N
4.6E-06 7.5E-07 3.lE.07 1.7E-07 1.3E-07 5.4E.08 1.5E-08 5.9E-09 3.0E-09 1.6E-09
.. :%I C."
7"t rl
- 1sa, B-5 I
MEAN RELATIVE DEPOSITION (m2)
ELEVATED RELEASE POINT - STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
SECTOR
.5 1.5 2.5 3.5 4.5 7.5
- 15.
- 25.
- 35.
45.
NNE 2.6E-10 3.OE-10 1.8E-10 1.3E-10 9.2E-11 5.7E-11 2.3E-11 1.3E-11 8.1E-12 5.8E-12 NE 1.9E-10 2.OE-10 1.2E.10 8.2E-11 6.1E-11 4.OE-11 1.6E-11 8.3E-12 6.OE-12 3.8E-12 ENE 1.4E-10 1.4E.10 8.7E-11 6.2E-11 4.5E-11 2.9E11 1.1E-11 5.7E-12 3.8E-12 2.6E-12 E
9.6E-11 9.5E-11 6.4E-11 4.6E-11 3.6E-11 2.3E-11 7.6E.12 3.7E-12 2.5E-12 8.5E-13 ESE 7.7E-11 1.OE-10 6.6E-11 4.8E 11 3.8E-11 2.3E-11 1.2E-11 5.2E-12 3.5E-12 2.OE-12 SE 2.3E-10 2.3E-10 1.4E-10 1.OE-10 7.5E-11 4.0E-11 1.7E-11 7.5E-12 4.8E-12 3.3E-12 SSE 4.2E-10 4.5E-10 2.6E-10 1.7E-10 1.6E-10 7.7E-11 3.3E-11 1.6E 11 l.lE-11 7.6E-12 S
6.4E-10 5.1E-10 3.OE-10 2.OE-10 1.5E-10 7.2E-11 2.9E-11 1.6E-11 1.1E-11 6.7E -12 SSW 3.OE-10 3.4E-10 1.'E-10 9.7E-11 7.2E-11 3.5E-11 1.3E.11 6.5E-12 4.9E-12 2.6E-12 SW 7.9E-11 2.1E-10 8.4E-11 5.OE-11 4.OE-11 2.1E-11 7.4E 12 3.8E-12 2.4E-12 1.8E-12 WSW 5.7E-11 2.3E10 1.)OE-10 6.2E-11 4.3E-11 2.3E-11 8.5E-12 4.3E-12 2.7E-12 1.8E-12 W
1.OE-10 3.4E-10 1.6E-10 9.8E-11 6.9E-11 2.9E-11 1.3E-11 6.2E-12 3.4E-12 2.OE-12 WNW 1.2E-10 4.1E-10 2.LE-10 1.3E-10 8.3E-11 3.9E-11 1.4E-11 7.OE-12 4.1E-12 2.6E-12 NW 1.2E.10 3.8E-10 2.LE-10 1.3E-10 8.2E-11 4.1E.11 1.7E-11 1.0E-11 6.3E 12 3.9E-12 NNW 2.3E.10 2.6E.10 3.OE-10 2.OE. 10 1.lE10 6.0E-ll 2.1E-ll l.lE-11 6.1E-12 3.9E-12 N
2.5E-10 3.7E-10 2.3E-10 1.5E-10 1.2E-10 7.1E-11 2.9E-11 1.7E-11 1.3E-11 5.2E.12 W171" From
.111, C117) 11-7V
- i z-- 11
- Ln C) 11
411-CD Z
ZZ4 B-6 I
MEAN RELATIVE DEPOSITION (m-2)
GROUND LEVEL RELEASE POINT-STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
SECTOR
.5 1.5 2.5 3.5 4.5 7.5
- 15.
- 25.
- 35.
45.
NNE 8.OE-09 1.2E-09 5.2E-10 3.1E-10 2.OE-10 9.9E-11 3.3E-11 1.6E-11 9.6E-12 6.OE-12 NE 5.1E-09 7.6E-10 3AE-10 2.OE-10 1.3E-10 6.9E-11 2.4E-11 1.1E-11 6.7E-12 4.1E-12 ENE 4.OE-09 6.1E-10 2.7E-10 1.6E-10 1.1E-10 4.8E-11 2.OE-11 7.6E-12 3.9E-12 2.5E-12 E
4.OE-09 6.1E-10 2.8E-10 1.6E-10
- 1. 1E-10 5.OE-11 1.8E-11 8.OE-12 4.2E.12 2.3E-12 ESE 5.3E-09 8.2E-10 3.5E-10 2.OE-10 1.4E-10 6.7E-11 2.1E-11 9.6E-12 5.6E-12 3.8E-12 SE 6.4E-09 9.6E-10 3.9E-10 2.4E-10 1.6E-10 7.1E-11 2.5E-11 l.lE-11 6.8E-12 4.1E-12 SSE 1.OE-08 1.5E-09 6.1E-10 3.5E-10 2.3E-10 1.1E-10 3.9E. 11 1.8E-11 1.1E-11 6.5E-12 S
8.7E.09 1.4E-09 5.8E-10 3.3E-10 2.3E-10 1.1E-10 4.OE-11 1.8E-11 1.OE-11 6.3E.12 SSW 3.7E-09 6.OE-10 2.6E-10 1.5E-10 1.OE-10 5.9E-11 1.5E-11 5.OE.12 3.OE-12 1.8E-12 SW 2.9E-09 4.4E-10 2.2E-10 1.2E-10 8.3E-11 3.5E-11 1.1E-11 3.OE-12 1.4E-12 8.7E-13 WSW 2.8E-09 4.6E-10 2.2E-10 1.3E-10 9.OE-11 3.7E-1 1.1E-11 4.2E-12 2.2E 12 1.4E.12 W
3.6E-09 5.9E-10 2.6E-10 1.5E-10 1.OE-10 4.6E-11 1.7E-11 6.9E-12 3.8E-12 2.2E-12 WNW 5.6E-09 8.7E-10 3.8E-10 2.3E-10 1.6E-10 7.3E-11 2.5E-11 1.OE-11 6.2E-12 3.8E-12 NW 1.OE-08 1.6E-09 6.8E-10 4.1E-10 2.7E-10 1.3E-10 4.5E-11 1.8E-11 1.1E-11 6.8E-12 NNW 1.lE-08 1.6E-09 6.9E 10 4.1E-10 2.8E-10 1.4E-10 5.2E-11 2.3E-11 1.3E-11 5.2E-12 N
1.2E-08 1.9E-09 8.IE-10 4.6E-10 3.2E-10 1.5E-10 5.8E-11 2.7E-11 1.7E-11 1.OE-11 X'-V I','
-44, B-7 i
ifoRMA"IM~o Oi"}N UR APPENDIX C ENVIRONMENTAL RADIATION MONITORING PROGRAM Appendix C contains the active environmental sampling stations for the Environmental Radiation Monitoring Program at Cooper Nuclear Station. Included in this appendix is a description of each sample and sample station along with maps showing the approximate location of each sampling station.
C-1
REMP SAMPLE STATION DESCRIPTION SAMPLE TYPES AND SAMPLE LOCATIONS (See Sample Station Locations Map - Figures C-1 and C-2)
Sample Station(a)
Sample Description - Type and Location No. 1 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
Outside the northwest edge of fence, east of the gate to the LLRW storage pad on the CNS site, NW1/4, S32, T5N, R16E, Nemaha County, Nebraska.
No. 2 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
On north side of county road access to the south portion of the CNS site, SW%, S32, T5N, R16E, Nemaha County, Nebraska.
No. 3 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
Located on the north side of the Brownville State Recreation Park access road near water gauging station, SE%4, S18, T5N, R16E, Nemaha County, Nebraska.
No. 4 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
Located Y2 mile south of Phelps City, Missouri, on west side of Highway "U," NE%4, S2, T64N, R42W, Atchison County, Missouri.
No. 5 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
One-fourth mile south and % mile east of Langdon, Missouri, on north side of road, west of railroad tracks, SW/4, S18, T64N, R41W, Atchison County, Missouri.
No. 6 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
One mile west of the end of Missouri State Highway "U," SW corner of the intersection, NW%, S34, T64N, R42W, Atchison County, Missouri.
C-2 62/06/02 1
Sample Station Sample Description - Type and Location No. 7 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
One-quarter mile east of Highway 67 at Nemaha on north side of road, SWA, S6, T4N, R16E, Nemaha County, Nebraska.
No. 8 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
One-half mile north, 3/4 mile west and 3 mile north of Nemaha on west side of road adjacent to the transmission line, NE%, S35, T5N, R15E, Nemaha County, Nebraska.
No. 9 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
Four miles north of Highway No. 136 on Highway No. 67. One mile east of Highway No. 67 and /2 mile north on west side of road, SWY4, S26, T6N, R15E, Nemaha County, Nebraska.
No. 10 Type: (1)
Air Particulate and Charcoal Filters (2)
Environmental Thermoluminescent Dosimetry Location:
One mile north of Barada, Nebraska, in SW corner of intersection, NE%, S14, T3N, R16E, Richardson County, Nebraska.
No. 11 Type: (1)
Water - Ground Location:
Plant well water supply header at well pits, NW%, S32, T5N, R16E, Nemaha County, Nebraska.
No. 12 Type: (1)
Water - River Location:
Sample (1) will be taken from the Missouri River immediately upstream from the Plant Intake Structure (River Mile 532.5). During periods when conditions warrant, Station 35 may be used as an alternate to Station 12 (upstream collection site) for sample type (1).
No. 20 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
On NNW boundary of NPPD property, east side of county road, SE%,
S30, T5N, R16E, Nemaha County, Nebraska.
C-3 2/06/02 1
Sample Station Sample Description - Type and Location No. 28 Type: (1)
Water - River (2)
Fish (3)
Sediment from Shoreline Location:
Samples (1) and (3) are taken from the Missouri River or its shore, downstream, near River Mile 530. Sample (2) is taken from the Missouri river % to 3 miles downstream from the plant site.
No. 35 Type: (1)
Fish (2)
Water - River (Alternate Site)
(3)
Food Products - Broadleaf Vegetation Location:
Sample (1) is taken from the Missouri River about one to three miles above CNS intake structure. During periods when conditions warrant, Station 35 may be used as an alternate to Station 12 (upstream collection site) for sample type (2). Samples (2) and (3) are taken about % mile south of the Brownville State Recreation Area in Sector A.
No. 44 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One-quarter mile south of Auburn Country Club on Highway No. 75,
%z mile east of Highway No. 75 at fence line north of county road, SE%, S27, T5N, R14E, Nemaha County, Nebraska.
No. 47 Type: (1)
Water - Ground Location:
At Falls City Municipal Water Supply Wells approximately 2 miles south of Rulo, Nebraska (out of Main Header Flow Meter), east side of road, SWA, S20, T1N, R18E, Richardson County, Nebraska.
No. 56 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One and one-fourth mile SW of Langdon, Missouri, on Highway "U",
on the right side of the highway, NW/4, S23, T64N, R42W, Atchison County, Missouri.
No. 58 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
Three miles south of Brownville, Nebraska, on county road, at the SE corner of the intersection, with the farm road leading to Sample Station No. 2, SE%, S31, T5N, R16E, Nemaha County, Nebraska.
No. 59 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile SSE of the CNS Elevated Release Point, in the vicinity of the levee at the south boundary of NPPD property, SE%, S32, T5N, R16E, Nemaha County, Nebraska.
C-4 2/06/02 1
Sample Station
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! i I M IN-" ILU Sample Description - Type and Location No. 61 No. 66 No. 67 No. 71 No. 79 No. 80 No. 81 No. 82 No. 83 Type: (1)
Location:
Type: (1)
Location:
Type: (1)
Location:
Type: (1)
Location:
Type: (1)
Location:
Type: (1)
Location:
Type: (1)
Location:
Type: (1)
Location:
Type: (1)
Location:
Milk - Nearest Producer One mile west of Brownville, NE, on Highway No. 136, then 1 mile north on county road, turn right and proceed approximately Y/zmile east on south side of road, NWY4, S1 3, T5N, RI 5E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two miles south of Nemaha, Nebraska, on Highway No. 67 - east side of highway, NW/4, S19, T4N, R16E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two miles west of Brownville, Nebraska, on Highway No. 136, then north 1%
/2miles on county road, and east 1/2 2mile, on north side of road, NE%, SI 1, T5N, RI 5E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two miles east of Phelps City, Missouri, on Highway No. 136, then south 11/2 Ymiles on county road, and west %/ mile, SE%, S6, T64N, R41W, Atchison County, Missouri.
Environmental Thermoluminescent Dosimetry One and 7/8 miles south of Brownville, Nebraska, on the east side of the paved road, NPPD property, SE1A, S30, T5N, R16E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two and 1/2/ miles south of Brownville, Nebraska, on the east side of the paved road, NPPD property, NE%, S31, T5N, R16E, Nemaha County, Nebraska.
Environmental Thermoluminescen't Dosimetry Two and 3/4/ miles south of Brownville, Nebraska, in the NE corner of the intersection of the paved county road and the CNS access road, NPPD property, NE%, S31, T5N, R16E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Seven eighths mile south of Cooper Nuclear Station in field, about '1/
mile east of farm buildings on NPPD property, SW%4, S32, T5N, R16E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two and %/ miles south of Nemaha, Nebraska, on Highway No. 67, then east one mile to the junction of the driveway and county road on the east side of the driveway, NEY4, S19, T4N, R16E, Nemaha County, Nebraska.
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C-5 2/06/02 1
Sample Station Sample Description - Type and Location No. 84 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
Two and % miles west of Brownville, Nebraska, on the south side of Highway No. 136, west of school, NW%, S22, T5N, R1SE, Nemaha County, Nebraska.
No. 85 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile east of Brownville, Nebraska, on Highway No. 136, then north % mile on the east side of the county road, NE%, S33, T65N, R42W, Atchison County, Missouri.
No. 86 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile west of Phelps City, Missouri, on Highway No. 136, then north 1 /z miles on Highway "D" - on the west side of road, SE'A, S22, T65N, R42W, Atchison County, Missouri.
No. 87 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile west of Phelps City, Missouri, on Highway No. 136, then south 1/2 mile on county road and % mile west on county road to the end of the road, NW%, S3, T64N, R42W, Atchison County, Missouri.
No. 88 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile west of Phelps City, Missouri, on Highway No. 136, then south 2 miles, west side of road, NWA, S11, T64N, R42W, Atchison County, Missouri.
No. 89 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
Two and %/ miles south of Phelps City, Missouri, on Highway "U",
then %/ mile west in the SE corner of the county road intersection, NE%, S14, T64N, R42W, Atchison County, Missouri.
No. 90 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One and %z miles west and % mile south of Langdon, Missouri, on Highway "U", then % mile west, SWI/4, S23, T64N, R42W, Atchison County, Missouri.
C-6 2/06/02 1
Sample Station Sample Description - Type and Location No. 91 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One half mile west of Rock Port, Missouri, on the south side of the intersection of Highway No. 136 and Highway No. 275 at the south side of the water tower, NW/4, S28, T65N, R41W, Atchison County, Missouri.
No. 94 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One quarter mile south of Langdon, Missouri, on the west side of the road, NE%, S24, T64N, R42W, Atchison County, Missouri.
No. 96 Type: (1)
Food Products - Broadleaf Vegetation Location:
Approximately 1 mile south of Brownville, Nebraska, along paved road in the road ditch in Sector R, SWA, S19, T5N, R16E, Nemaha County, Nebraska.
No. 99 Type: (1)
Milk (Other Producer)
Location:
One and % miles south of Shubert, Nebraska, on the west side of Highway No. 67, NE%, S24, T3N, R15E, Richardson County, Nebraska.
No. 100 Type: (1)
Milk (Other Producer)
Location:
Two miles south and 1 mile west of Shubert, Nebraska, SW%, S23, T3N, R15E, Richardson County, Nebraska.
No. 101 Type: (1)
Food Products - Broadleaf Vegetation Location:
Five and t/2 miles east and Y/2 mile north of Rock Port, Missouri, near the junction of Highway No. 136 and Highway No. 59, in Sector D.
Encompasses portions of several sections, T65N, R40W, Atchison County, Missouri.
NOTES:
(a)
Sample Station numbers missing from the sequence are for inactive or discontinued Sample Stations.
C-7 2/06/02
NEBRASKA MISSOURI NEBRASKA PUBUC POWER DISoIwC COOPER NUCLEAR sTANON Site Sampling Station Location Map FIGURE C-I C-8 8/26/99
4 A,
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SCALE IN MILES cAO FILE: uMoos3730 C-9 8/26/99