ML021280166
| ML021280166 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 02/06/2002 |
| From: | Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML021280166 (192) | |
Text
INFORMATION ONLY Assessment Manual
-ODAM For assessment of Gaseous and Liquid Effluents at COOPER NUCLEAR STATION Brownville, Nebraska INFORMATION ONLY 28872.WPD
COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL
-ODAM LIST OF EFFECTIVE PAGES Page Page Rev/Date 8/26/99 8/26/99 8/26/99 8/26/99 8/26/99 8/26/99 8126/99 8/26/99 8/26/99 8/26/99 2/22/01 8/26/99 2/22/01 8/26/99 8/26/99 8/26/99 8/26/99 ii 1
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Page 1 of 2 Appendix A:
(Deleted) 1/27/89 Appendix B:
B-1 B-2 B-3 B-4 B-5 B-6 B-7 Original Original Original Original Original Original Original C-7 C-8 C-9 Appendix D:
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COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL
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COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL FOR GASEOUS AND LIQUID EFFLUENTS Page 1.0 Introduction..........................................................
-1 2.0 Liquid Effl uent.......................................................
-2 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 Projected Personal Maximum Dose.............................. 3.0 Gasseous Effluent.....................................................
3.1 Introduction................................................
3.2 Radioactivity in Gaseous Effluent...............................
3.3 Main Condenser Air Ejector 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..............................
-1 8/.
-17
-17
-17
-19
-20
-20
-24
-27
-29
-30
-31
-34 26/99
"-i-
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 Projected Organ Dose Due to Gaseous Effluent....................
-37 3.10 Dose Rate Due to Tritium, Iodines, and Particulates in Gaseous Effluents-38 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
-ii-
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.
10/15/96 4--
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 10i8 pCi/ml, 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 F1 c=c ZI '
2 where Ci
=
concentration of radionuclide i
in liquid radwaste released (gCi/ml)
C
=
concentration of radionuclide i in the receiving stream (iCi/ml)
F1
=
release rate of liquid radwaste (ml/sec)*
F2
=
dilution flow of receiving stream of water (ml/sec)*
- F1, 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 radionucide i in a receiving stream is very conservative as it is based on not exceeding the concentration limits in 10CFR20 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:
- CqciVrq Vdq where Czi Cqci Vrq Vdq
=
concentration of radionuclide i in the receiving stream (1tci/ml)
=
concentration of radionuclide i in the quarterly composite sample (pci/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)
For all other nuclides Czi = Cmci VrM Vdm where Czi Cmci Vrm Vdm
=
concentration of radionuclide i in the receiving stream (Ici/ml) concentration of radionuclide i in the monthly composite sample (Rci/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 P3-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 gCi/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 1-131 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 FMPCbP =
- CbpI identified I MPC.
10/15/96 where FMPCbp Cbpi MPCi 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 1-131 and principal gamma emitters) in batch sample taken prior to release ([Ci/ml).
effluent concentration beyond the site and exclusion area boundary of radionucide i per 10 CFR Part 20, Appendix B, Table 2, Column 2 (gCi/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 1-131 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 A
FS2 g + Bkg FMPCb Fs1 8/26/99 I
where S
=
radiation monitor alarm setpoint (cpm or gCi/ml)
A
=
counting rate (cpm/ml) or activity concentration (gCi/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 gCi/ml or gCi/ml per gCilml)
Fs,
=
maximum flow in the batch release line (gal/min)**
Fs2
=
minimum flow in the discharge canal (gal/min)**
Bkg
=
monitoring instrument background (cpm or gCi/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 10-8 where Cbp gross or total radioactivity concentration in batch sample taken prior to release (iCi/ml)
I x 10-8
=
effluent concentration beyond the site and exclusion area boundary of unidentified radionuclides (ILCi/ml)
- A equals C.
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 FMPCb, 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,
= x C
identified 1
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).
C=
concentration of radionuclide i (including 1-131 and principal gamma emitters) in weekly composite sample(s)
([+/-Ci/ml)
When FMPCc 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 = FMPC,, - E' 10/15/96 where E=
(Quarterly average fraction of the effluent concentration in the discharge canal due to 1-131 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
F S=
-A'g
+ Bkg FMPCC FSI where A
=
counting rate (cpm/ml) or activity concentration (gCi/ml) of weekly composite sample in the laboratory.
Terms g, Fs8, 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:
c, FMPC X cc C
1 x10-8 8/26/99 I
where C,
=
gross or total radioactivity concentration in continuous aqueous release (gCi/ml) 1 x 108
=
effluent concentration beyond the site and exclusion area boundary of unidentified radionuclides ([Ci/ml)
The value of FMPC, 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.
8/26/99 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 E
1 E Oi 3785(TE-TB) k F2 k
- 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)
O
=
quantity of radionuclide i represented by sample k which is released as an effluent within the time boundaries TB and TE (MCi) 8/26/99 I
MPC, =
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.
- Wth 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.
0 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:
IAD k = 3.785 X 10-3 e Aeani C ik -At4k
)
2k Dan AD Dn k ank where ADak =
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 12/29/88 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).
Aani
=
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 - mini gal 3.785 x 103- = 3785 ml/gal x 10Ci/gCi C
=
the concentration of radionuclide i in the undiluted liquid waste to be discharged (jiCi/ml), i.e., in the sample k Atk
=
elapsed time of release represented by sample k during which radionuclide i is discharged at concentration Ck, i.e., the duration of the release represented by sample k (minutes)
(F/1/F2) =
the quotient of the release flow, F 1, and the dilution flow, F 2, during the release represented by sample k 12/29/88 Pathway-to-dose transfer factors, Aka,, 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 = 5F,. 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 = 5F.. Variables F1, 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 05/16/94 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 91 a,=x-an where P.
91 X
Dan projected dose commitment (mrem) to organ n (including total body of age group a during the current quarter)
=
number of days in a quarter
=
number of days to date in current quarter
=
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 I
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.
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 D3.2.3-1. 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 104_N -
F
- g.
where Q
=
total radioactive noble gas release via a gaseous effluent stream during a given time interval (CCi)
N
=
net counts accumulated during the time interval g
=
effluent noble gas monitor counting rate response cpm F
=
gaseous effluent stream discharge rate (cfm) 2.8 x 10' = conversion constant (cm 3/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 -h
+ Bkg F
where S
=
the alarm setpoint (mr/hr) h
=
monitor response to activity concentration of SJAE offgas being monitored (mr/hr per (ItCi/cm3))
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 1.OE6pCi 1 ft 3
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.l.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 radionucides 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 S=.0
-f-C.
DF*) + Bkg 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 Cq h.P i
S= 1.06 f-
+ Bkg
ýC DF")
i where S
=
the alarm setpoint (cpm, mr/hr, or jiCi/cm3) h
=
monitor response to activity concentration of effluent being monitored, (cpm per [Ci/cm3, mr/hr per jCi/cm3, or gCi/cm3 per gCi/cm3).
Ci relative concentration of noble gas radionuclide i in effluent at the point of monitoring (*ICi/cm3)
X/Q
=
atmospheric dispersion from point of ground-level or split-wake release to the location of potential exposure (sec/m3)
S DFi factor converting elevated release rate of radionuclide i to total I body dose equivalent rate at the location of potential exposure (mrem)/(yr.yCilsec) 8/26/99 V
DFi
=
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./Ci/m 8 )
f
=
flow of gaseous effluent stream, i.e., flow past the monitor (ft3/min)
Bkg
=
monitoring instrument background (cpm, mr/hr, or liCi/cm3) 1.06
=
500 mrem/yr x 60 sec/min x 35.3 ft3/m3 x 1.0 m3/1.0 x 106 cm 3 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, DFi' depends upon local 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 I
ERP Elevated W
1,800 8.2 x 10"8 The applicable dose conversion factors, DFPi and DFXi 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=**
Q A
+
(Q.V X) yJ" where D = noble gas gamma dose to air (mrad)
QC. = Z AQS.= cumulative release of noble gas nuclide i from stack (tCi).
1 1
Aye
=
factor converting unit noble gas stack release to ground level air dose from overhead plume gamma radiation (mrad4LCi).
Ay,
=
factor converting time integrated, ground level concentration of noble gas to air dose from gamma radiation mrad IiCi sec m )
8/26/99 1
=Ze AQcvi
=
cumulative release of noble gas nuclide i from building time vents (gCi).
(XW 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 Ay¥*
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 D=
(Q3
+
Q X~3
=i
+
Q*cvi X
Where D
=
noble gas beta dose to air (mrad)
(X/Q). =
long-term average atmospheric dispersion factor for stack releases (sec/mi)
Api
=
factor converting time integrated ground level concentration of noble gas radionuclide i to air dose from beta radiation mrad (pCi seC)/m 3 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' sec/rm at the NNW site boundary (X/Q), =
3.4 x 10 6 sec/mrn Beta radiation-to-air dose conversion factors, Api, 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, Fa, 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 having half-lives greater than eight days. Noble gases not considered.
8/26/99
A j CikFaIAtJ Qk j Ci j
7 where Qik Cik FAt
=
the quantity of radionucide i released in a given effluent stream based on analysis k (Ci) concentration of radionucide i in gaseous effluent identified by analysis k (Cinm) or (ICi/cm3 )
=
effluent stream discharge rate during time increment At,(m 3/sec)
=
elapsed time in increment j during which radionuclide i at concentration Ck is being discharged (sec) 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, Qk, 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
[
3.7.2 Alternate Method Alternatively, the dose commitment to a person beyond the site and 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:
D., = QjiTA ~Xd)
+
Tj ean lQbse D~
QQ e iQ release via a vent:
am'Q A.
e Q) e,,IDl where Dawk
=
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)
TA..j factor converting airborne concentration of radionuclide i to dose commitment to organ n of a person in age group a
( mrem '
(Ci sec)/W)
TG. =
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/m2 )
(D/Q) =
relative deposition factor (min) 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
= w D,*
+ mD., +
Dauq m
q and the dose accumulated as a result of a vent discharge is D
=
+ m Danvm +
anD q
Doses committed during the same time period due to discharges from the stack and vents are additive, thus:
D. =Da +D D.
V where D.,
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
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, Aeam, 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/m3
(
= 4.6 x 10-10 M-2 (Xd)
= 4.4 x 1O-7 sec/m 3 (D)
= 9.5 x 10-10 M-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)
=1.2x10'° M -2 (Q)
=3.7 x 1 1 n
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 Q3 i QkS p for a vent release XE E
Dakv -,
p TAanjp 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.
8/26/99 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= (
y,
+
P<
=
a i Qc*PI Q
), --
P"P, 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/ICi)
Py,
=
factor converting time integrated, ground level concentration of noble gas nuclide i to air dose from gamma radiation
(
.mrem '
ptCi sec m3 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 Pycsi and Pyv1 for.
this receptor, 1.1 miles west of the station, appear in Table 3-5.
8/26/99 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 i Qc Qi+Q i)
-sp, where Do
=
noble gas beta dose to skin (mrem)
Si
=
factor converting time integrated ground level concentration of noble gas radionuclide i to skin dose from beta radiation mrem
([tCi sec)/m3 Values of SRi 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 Projected 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=
3 1 D where:
X PD 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 1
3.10 Dose Rate Due to Tritium, Iodines, 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 TE-TB k i Q aa ( TA.
Q) earn For effluent from a ground-level release point, i.e., a building vent, the equation is:
10-6 E
D. -_
Q,.
( X)TA.
TE-TB k i Q) a Dose rates from separate release points may be combined to give D. =D..
+ E Dam V
where Dan
=
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 (mrem/hr) 8/26/99
Da,,Danv TE TB TE-TB (Q)T,(X/Q)v TA.13i 10-6
=
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
(
mrem (Ci sec)/m')
-conversion, 10'Ci/,uCi 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 I
4.0 Dose Commitment From Releases Over Extended Time 4.1 Releases Durine 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 P as in
§ 3.6.
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
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 ordinarily assumed 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 thepublic 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.
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 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
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 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.
8/26/99
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 1
bumy w Inoulml AUMOPMn (2) orywo A I C N A 0 C 0 Vogt Flow EVIS
-Vmt Babods Im EVIS Radwam saft Fad" TW"Bdft Nodwasto Bwft mnlaft hum" CootaMtBd Am Figin 3-1 Qý UHM S&M TMUM BW MODIUM
. -48.
2/22/01 pw=to*MW woefrdwprmwte*mw Chw" bM=mt TaMe M22 I ý WoMmU nuotodwflowommut
TO DISCHARGE FIGURE 3.14A - LIQUID EFFLUENT STREAM, TREATMENT AND MONITORINGAND 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 Bullding 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 1
96" x 48" 15 73405 3.82 70 90 Partial Elevated 4
48" x 96" 1.3 50710(a) 26.4 70 90 Ground Level 1
72" I.D.
Horizontal discharge at rooftop 67070 39.5 70 90 Ground Level 1
14" 1.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.
2/22/01 Table 3-2 Computed Release of Radioactive Noble Gases In Gaseous Effluent From Cooper Nuclear Station Stack Release Plant Vents Release Nuclide (Cilyr)
Fraction (Ci/yr)
Fraction Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 Kr-89 Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Total 3.60E+01 6.50E+01 2.OOE+02 2.13E+02 2.13E+02 1.OOE+03 3.OOE 00 1.51E+02 7.20E+01 2.64E+02 1.20E+03 8.77E+02 4294.
8.38E-03 1.51E-02 4.66E-02 4.96E-02 4.96E-02 2.33E-01 6.99E-04 3.52E-02 1.68E-02 6.15E-02 2.79E-01 2.04E-01 1.0 0
7.10E+01 0
1.33E+02 2.33E+02 0
0 2.63E+03 6.96E+02 1.06E+03 0
1.41E+03 6233.
0 1.14E-02 0
2.13E-02 3.74E-02 0
0 4.22E-01 1.12E-01 1.70E-01 0
2.26E-01 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 DFES for Stack Release5 (mrem)/(yr.pCi/sec)5 mrem/gCi5 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 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 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
'Based on reference meteorology; applicable at the site boundary, ERP.
Factor D.Fifor Ground-Level or Split-Wake Release (mrem)/(yr.)uCi/m')
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 1,250 meters NNW of the 8/26/99 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 AN~
mrad mrad mrad g-C gCi sec/m 3 pCi sec/m 3 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 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 aDose at NNW site boundary 8/26/99 I
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 Py vi Spi mrem mrem mrem iiCi iCi sec/rmn 1Ci sec/mr 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 aReceptor located at 1.1 miles west of Station bBased on reference meteorology at Cooper Station 8/26/99 I
APPENDIX A (DELETED) 1/27/89 A-1
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-1
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 4.1E-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 1.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.OE-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 5.0E-1O 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 B-2 I
.1 IF'
UNDEPLETED MEAN RELATIVE CONCENTRATION (sec/mr3)
GROUND LEVEL RELEASE POINT - STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
B-3 i
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 3.1E-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 6.1E-08 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 8.1E-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 1.1E-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 2.4E-08 1.1E-08 6.1E-09 3.5E-09
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 1.1E-08 9.OE-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-09 5.6E-10 SE 8.1E-09 1.9E-08 1.6E-08 1.3E-08 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.OE-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 7.8E-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 B-4I
.1 9I'
DEPLETED MEAN RELATIVE CONCENTRATION (sec/m 3)
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 1.1E-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 3.1E-07 1.7E-07 1.3E-07 5.4E-08 1.5E-08 5.9E-09 3.OE-09 1.6E-09 B-5 I
- o. 1.,
MEAN RELATIVE DEPOSITION (m"2)
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 2.9E-11 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.OE-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 1.1E-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 1.4E-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 2.3E-16 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 2.1E-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 2.1E-10 1.3E-10 8.2E-11 4.1E-11 1.7E-11 1.OE-11 6.3E-12 3.9E-12 NNW 2.3E-10 2.6E-10 3.OE-10 2.OE-10 1.1E-10 6.OE-11 2.1E-11 1.1E-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 B-6
.4 I'l
MEAN RELATIVE DEPOSITION (m")
GROUND LEVEL RELEASE POINT - STANDARD DISTANCES COOPER NUCLEAR STATION NEBRASKA PUBLIC POWER DISTRICT DISTANCE (miles)
B-7 i
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 3.4E-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 1.1E-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-1Q 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-11 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 1.1E-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 8,1E-10 4.6E-10 3.2E-10 1.5E-10 5.8E-11 2.7E-11 1.7E-11 1.OE-11
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, NW%, 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, SW1/4, 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, SE1/44, 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%, 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, SWV4, 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 comer of the intersection, NWY4, S34, T64N, R42W, Atchison County, Missouri.
2/06/02 1
C-2
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 3/4 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, SW%,
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 comer of intersection, NE%, S14, T3N, R16E, Richardson County, Nebraska.
No. 11 Type: (1)
Water - Ground Location:
Plant well water supply header at well pits, NW1/44, 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
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 1/2 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, 1/2 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, SW1/4, S20, TIN, 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, 823, 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 comer of the intersection, with the farm road leading to Sample Station No. 2, SE1/4, 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.
2/06/02 1
C-4
Sample Description - Type and Location Sample Station No. 61 No. 66 No. 67 No. 71 No. 79 No. 80 No. 81 No. 82 No. 83 2/06/02 1
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 2 mile east on south side of road, NW%, S13, T5N, R1 5E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two miles south of Nemaha, Nebraska, on Highway No. 67 - east side of highway, NW1A, S19, T4N, R16E Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two miles west of Brownville, Nebraska, on Highway No. 136, then north 1% miles on county road, and east Y2 mile, on north side of road, NE1/4, S11, T5N, R1 5E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two miles east of Phelps City, Missouri, on Highway No. 136, then south 1% miles on county road, and west % mile, SEY4, 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, SE%, S30, T5N, R16E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two and 1/2a miles south of Brownville, Nebraska, on the east side of the paved road, NPPD property, NEY4, S31, T5N, R16E, Nemaha County, Nebraska.
Environmental Thermoluminescent Dosimetry Two and 3 miles south of Brownville, Nebraska, in the NE comer 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, SWY4, 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, NE%, S19, T4N, R16E, Nemaha County, Nebraska.
I I I C-5
Sample Station Sample Description - Type and Location No. 84 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
Two and 2 miles west of Brownville, Nebraska, on the south side of Highway No. 136, west of school, NWY4, S22, T5N, R15E, Nemaha County, Nebraska.
No. 85 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile east of Brownville, Nebraska, on Highway No. 136, then I
north / mile on the east side of the county road, NE%, S33, T65N, I
R42W, Atchison County, Missouri.
I No. 86 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile west of Phelps City, Missouri, on Highway No. 136, then I
north 11/2 miles on Highway "D" - on the west side of road, SE%, S22, I
T65N, R42W, Atchison County, Missouri.
I No. 87 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile west of Phelps City, Missouri, on Highway No. 136, then t
south /2 mile on county road and % mile west on county road to the I
end of the road, NW1/44, S3, T64N, R42W, Atchison County, Missouri.
I No. 88 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One mile west of Phelps City, Missouri, on Highway No. 136, then I
south 2 miles, west side of road, NW1/4, Si1, T64N, R42W, Atchison County, Missouri.
No. 89 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
Two and 2 miles south of Phelps City, Missouri, on Highway "U",
then 2 mile west in the SE comer of the county road intersection, NE1/4, S14, T64N, R42W, Atchison County, Missouri.
No. 90 Type: (1)
Environmental Thermoluminescent Dosimetry Location:
One and % miles west and % mile south of Langdon, Missouri, on Highway "U", then / mile west, SW,4, $23, T64N, R42W, Atchison County, Missouri.
2/06/02 1
C-6
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, NW1/44, 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, NE1/4, 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, SW1M4, 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, SW4, S23, T3N, R15E, Richardson County, Nebraska.
No. 101 Type: (1)
Food Products - Broadleaf Vegetation Location:
Five and 1/2 miles east and 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.
2106/02 1
C-7
S MISSOURI 8/26/99 NEBRASKA NEBRASKA PUBLIC POWEt DISTICT COOPER NUCLEAR STATION Site Sampling Station Location Map FIGURE C-i N
C-8
I I
n 4D
APPENDIX D ODAM SPECIFICATIONS
OFFSITE DOSE ASSESSMENT MANUAL APPENDIX D TABLE OF CONTENTS D 1.0 Use and Application.............................................
D 1.0-1 D 2.0 Definitions....................................................
D 2.0-1 D 3.0 ODAM Specifications Applicability...............................
D 3.0-1 D 3.1 LIQUID EFFLUENTS D 3.1.1 Liquid Effluents Concentration................................
D 3.1-1 D 3.1.2 Liquid Waste Concentration..............................
D 3.1-5 D 3.1.3 Liquid Effluents Dose.....................................
D 3.1-6 D 3.1.4 Outside Temporary Storage of Radioactive Liquid................
D 3.1-9 D 3.2 GASEOUS EFFLUENTS D 3.2.1 Gaseous Effluents Concentration.............................
D 3.2-1 D 3.2.2 Noble Gas Dose......................................
D 3.2-3 D 3.2.3 Iodine and Particulates....................................
D 3.2-5 D 3.2.4 Offgas Treatment System.................................
D 3.2-10 D 3.2.5 Exhaust Ventilation Treatment System.......................
D 3.2-12 D 3.2.6 Hydrogen Concentration..................................
D 3.2-14 D 3.2.7 Primary Containment Venting and Purging....................
D 3.2-15 D 3.3 INSTRUMENTATION D 3.3.1 Liquid Effluent Monitoring..................................
D 3.3-1 D 3.3.2 Gaseous Effluent Monitoring.................................
D 3.3-8 D 3.4 LIQUID/GASEOUS DOSE D 3.4.1 Liquid/Gaseous Effluents Dose..............................
D 3.4-1 D 3.5 SOLID RADIOACTIVE WASTE D 3.5.1 Solid Radioactive Waste...................................
D 3.5-1 D 4.0 MONITORING PROGRAM D 4.1 Monitoring Program Compliance..............................
D 4.1-1 D 4.2 Monitoring Program Concentration............................
D 4.2-1 D 4.3 Monitoring Program Dose................................
D 4.3-1 D 5.0 MISCELLANOUS PROGRAMS/REPORTS D 5.1 Interlaboratory Comparison Program.........................
D 5.1-1 D 5.2 Annual Radiological Environmental Report......................
D 5.2-1 D 5.3 Radioactive Effluent Release Report...........................
D 5.3-1 D 5.4 Special Reports..........................................
D 5.4-1 D 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid).........................
Use and Application D 1.0 D 1.0 USE AND APPLICATION The Offsite Dose Assessment Manual (ODAM) Specifications are contained in Section 3.0 of this appendix.
They contain operational requirements, Surveillance Requirements, and reporting requirements.
Additionally, the Required Actions and associated Completion Times for degraded Conditions are specified.
The format is consistent with the Technical Specifications (Appendix A to the CNS Operating License).
The definitions contained in Technical Specifications Section 1.1, "Definitions," apply to the ODAM Specifications.
Defined terms are shown in all capital letters, consistent with the Technical Specifications.
The rules of usage for the ODAM Specifications are the same as those for the Technical Specifications.
These rules are found in Technical Specifications Sections 1.2, "Logical Connectors;" 1.3, Completion Times;" and 1.4; "Frequency."
Figure D2.a-1 D 2.0 DEFINITIONS Figure D2.a-1 (Page I of 1)
Site and Exclusion Area Boundaries CNS ODAM D 2.0-1 12/20/01
ODAM Specifications Applicability D3.0 D 3.0 ODAM Specifications Applicability DLCO 3.0.1 DLCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in DLCO 3.0.2.
DLCO 3.0.2 Upon discovery of a failure to meet a DLCO, the Required Actions of the associated Conditions shall be met, except as provided in DLCO 3.0.5.
If the DLCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated.
DLCO 3.0.3 When a DLCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, action shall be initiated to:
- a.
Restore compliance with the DLCO or associated ACTIONS, and
- b.
Enter the circumstances into the Corrective Action Program.
-- -- NOTE ---
DLCO 3.0.3.b shall be completed if DLCO 3.0.3 is entered.
Exceptions to this Specification are stated in the individual Specifications.
DLCO 3.0.4 Not Applicable to ODAM Specifications.
DLCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to DLCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.
ODAM Specifications Applicability D3.0 D 3.0 ODAM Specifications Applicability DSR 3.0.1 DSR 3.0.2 DSR 3.0.3 DSRs shall be met during the MODES or other specified conditions in the Applicability for individual DLCOs, unless otherwise stated in the DSR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the DLCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the DLCO except as provided in DSR 3.0.3.
Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
The specified Frequency for each DSR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance on a "once per..." basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the DLCO not met may be delayed, from the time of discovery, up to-24 hours or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance of the Surveillance.
If the Surveillance is not performed within the delay period, the DLCO must immediately be declared not met, and the applicable Condition(s) must be entered.
When the Surveillance is performed within the delay period and the Surveillance is not met, the DLCO must immediately be declared not met, and the applicable Condition(s) must be entered.
Liquid Effluents Concentration D 3.1.1 D 3.1 LIQUID EFFLUENTS D 3.1.1 Liquid Effluents Concentration DLCO 3.1.1 APPLICABILITY:
The concentration of radioactive material in water beyond the Site and Exclusion Area Boundary (Figure D2.a-1) due to radioactive liquid effluent shall not exceed:
- a.
The concentration specified in 10 CFR Part 20.1302 for radionuclides other than dissolved or entrained noble gases; and
- b.
2 x 10-4 pCi/ml total activity concentration for dissolved or entrained noble gases.
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A.1 Initiate action to Immediately radioactive material restore concentration beyond the Site and to within limits.
Exclusion Area Boundary due to radioactive liquid effluent exceeds limits.
Liquid Effluents Concentration D 3.1.1 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.1.1 Perform radioactive liquid waste sampling In accordance and activity analysis.
with Table D3.1.1-1 DSR 3.1.1.2 The analytical results shall be used with In accordance methods in the ODAM to verify that the with ODAM average concentration beyond the Site and Section 2.4 Exclusion Area Boundary does not exceed DLCO 3.1.1 when SR-89, SR-90, and Fe-55 concentrations are averaged over no more than 3 months and other radionuclide concentrations are averaged over no more than 31 days.
Liquid Effluents Concentration D 3.1.1 Table D3.1.1-1 (Page 1 of 2)
Radioactive Liquid Waste Sampling and Analysis LIQUID SAMPLE LOWER RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE LIMIT OF DETECTION TYPE TYPE FREQUENCY FREQUENCY ANALYSIS (LLD)(h)
Batch Waste Grab Each batch (a)
Each batch (a)
Principal Gamma 5 X 10 pCi/ml (i)
Release Tanks (c) sample Emmitters 0)(k) 1-131 1 X 10 pCi/ml Grab sample One batch/
31 days (b)
Dissolved and 1 X lOs pCi/ml 31 days (a)
Entrained Gases (gamma emitters)
Proportional Each batch (a) 31 days (b)
H-3 I X lOs pCi/mI Composite of grab samples (f)
Gross Alpha I X lOF pCi/ml 92 days (b)
Sr-89 5 X 10W pCi/ml Sr-90 5 X 10'pCi/ml Fe-55 1 X 1O6 pCi/ml Plant Service Grab Sample 7 days 7 days (b)
Principal Gamma 5 X IO7 pCi/mI (i)
Water Effluent (d)
Emitters (j(k)
Plant Continuous Proportional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7 days (b)
Principal Gamma 5 X lO iCi/ml (i)
Discharge (e)
Composite of Emitters (j) (k)
Grab Samples (g) 1-131 1 X 104 pCi/ml Grab Sample 31 days 31 days (b)
Dissolved and I X 1Os pCi/ml Entrained Gases (gamma emitters)
Proportional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 31 days (b)
H-3 1 X 1V pCi/ml Composite of Grab Samples (g)
Gross Alpha 1 X lOf ICi/ml 92 days (b)
Sr-89 5 X 104 pCi/ml Sr-90 5 X 10W pCi/ml Fe-55 1 X lO6 pCi/iml (a)
Complete prior to each release.
(b)
Analysis may be performed after release.
(c)
A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly mixed.
(d)
A grab sample of plant service water effluent shall be analyzed at least once each week in accordance with Table D3.1.1-1, Plant Service Water Effluent. In the event the radioactivity concentration in a sample exceeds 3 X 10-6 pCi/dm, or in the event the plant service water effluent monitor indicates the presence of an activity concentration greater than 3 X 106 PCi/ml, sampling and analysis according to Table D3.1.1-1, Plant Continuous Discharge, shall commence and shall be performed as long as the condition persists.
(e)
A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the continuous release.
Table D3. 1.1-1 (Page 2 of 2)
Radioactive Liquid Waste Sampling and Analysis (f)
A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.
(g) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, daily grab samples shall be collected in proportion to the rate of flow of the effluent stream. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.
(h)
The LLD is the smallest concentration of the radioactive material in a sample that will be detected with 95% probability (5%
probability of falsely concluding that a blank observation represents a "real* signal).
For a particular measurement system (which may include radiochemical separation):
LLD =
(4"66)(Sb)
(E)(V)(222) (Y) e At Where:
LLD is the "a priori" lower limit of detection as described above (as picocurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation),
V is the sample size (in units of mass or volume),
2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and
&t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).
The value of s b used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. Typical values of E, V, Y and At shall be used in the calculation.
(i)
For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD may be increased inversely proportionally to the magnitude of the gamma yield (i.e., 5 X 107/1, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the value specified in 10 CFR 20, Appendix B, Table 2, Column 2.
])
The principal gamma emmitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analysis should not be reported as being present at the LLD level. When unusual circumstances result in LLD's higher than required, the reasons shall be documented in the Radioactive Effluent Release Report.
(k) If an isotopic analysis is unavailable, batch releases may be made for up to 14 days provided the gross beta/gamma concentration to the unrestricted area is s 1 x 10" pc/ml and the sample is analyzed when the instrumentation is once again available.
Liquid Waste Concentration D 3.1.2 D 3.1 LIQUID EFFLUENTS D 3.1.2 Liquid Waste Concentration DLCO 3.1.2 The concentration of radioactive materials in liquid wastes from pre-release analysis shall be __ 0.01 pCi/ml, excluding tritium and noble gases.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Concentration of A.1 Appropriate parts of the liquid Prior to liquid waste radioactive materials in radwaste treatment system discharge liquid wastes from pre-shall be used to reduce the release analysis concentration.
> 0.01 pCi/ml, excluding tritium and noble gases.
B.
Required Action and B.1 Prepare and submit a Special 31 days following the associated Completion Report to the NRC pursuant to end of the quarter in Time not met.
Specification D 5.4 that which the limit was identifies equipment or exceeded AND subsystems not OPERABLE and the reason for the Radioactive liquid waste inoperability, action(s) taken to being discharged without restore the inoperable treatment in excess of equipment to OPERABLE 0.01 pCi/mI, excluding status and a summary tritium and noble gases.
description of the action(s) taken to prevent a recurrence.
D 3.1-5 10/10/01
Liquid Effluents Dose D 3.1.3 D 3.1 LIQUID EFFLUENTS D 3.1.3 Liquid Effluents Dose DLCO 3.1.3 The dose to a Member of the Public due to radioactive material in liquid effluents beyond the Site and Exclusion Area Boundary (Figure D2.a-1) shall be limited to:
- a.
_ 1.5 mrem to the total body or 5 5.0 mrem to any body organ during any calender quarter; and
- b.
s 3.0 mrem to the total body or s 10.0 mrem to any body organ during any calender year.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Calculated dose due to A.1 Prepare and submit a Special 31 days following the radioactive material in Report, in lieu of any other end of the quarter in liquid effluents beyond report, pursuant to Specification which the limit was the Site and Exclusion D 5.4 to the NRC which exceeded Area Boundary identifies the cause(s) for exceeds the limit, exceeding the limit(s) and defines the corrective actions to be taken.
D 3.1-6 10/10/01
Liquid Effluents Dose D 3.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
Calculated dose due to B.1 Prepare and submit a Special 31 days radioactive material in Report, in lieu of any other liquid effluents beyond report, pursuant to Specification the Site and Exclusion D 5.4 to the NRC which: 1)
Area Boundary defines actions to be taken to exceeds two times the reduce releases and prevent
- limit, recurrence, and 2) results of an exposure analysis including effluent pathways and direct radiation to determine whether the dose or dose commitment to a Member of the Public due to radiation and radioactive releases from Cooper Station during the calender year through the period covered by the calculation was __ 75 mrem to the thyroid and __ 25 mrem to the total body and all other body organs.
Liquid Effluents Dose D 3.1.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.3.1 Perform an assessment of compliance with 31 days DLCO 3.1.3.
DSR 3.1.3.2 Project a prospect of compliance with In any quarter in DLCO 3.1.3 for radioactive liquid releases without which Radioactive radwaste system in operation.
liquid releases are made and the radwaste system is not operated CNS ODAM D 3.1-8 10/10/01 1
Outside Temporary Storage of Radioactive Liquid D 3.1.4 D 3.1 LIQUID EFFLUENTS D 3.1.4 Outside Temporary Storage of Radioactive Liquid DLCO 3.1.4 Radioactive liquid contained in unprotected outdoor temporary liquid storage tanks shall conform to the requirements of Technical Specification (TS) 5.5.8.b.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Level of radioactivity A.1 Suspend addition of Immediately exceeds the limits of TS radioactive material.
5.5.8.b.
AND A.2 Begin measures to reduce Immediately content to within the limits of TS 5.5.8.b.
AND A.3 Describe the events leading to Prior to submittal of the condition in the next Radioactive Radioactive Effluent Release Effluent Release Report.
D 3.1-9 10/10/01
Outside Temporary Storage of Radioactive Liquid D 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.4.1 Sample and analyze radioactive liquid located in 7 days during unprotected outdoor temporary liquid storage tanks addition of for level of radioactivity.
radioactive liquid to the tanks 10/10/01 1
Gaseous Effluents Concentration D 3.2.1 D 3.2 GASEOUS EFFLUENTS D 3.2.1 Gaseous Effluents Concentration DLCO 3.2.1 The dose rate beyond the Site and Exclusion Area Boundary (Figure D2.a-1) due to radioactive gaseous effluents shall be limited to the following:
- a.
For noble gases, s 500 mrem per year to the total body and
_s 3000 mrem per year to the skin; and
- b.
For H-3, 1-131, 1-133, and radioactive material in particulate form with half lives - 8 days, _* 1500 mrem per year to any organ when;
- 1.
The dose rate due to H-3, Sr-89, Sr-90, and alpha emitting radionuclides is averaged over _ 3 months and;
- 2.
The dose rate due to other radionuclides is averaged over :_ 31 days.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Dose rates beyond the A.1 Decrease release rate to Immediately Site and Exclusion Area comply with the limits.
Boundary due to radioactive gaseous effluents exceeds limits.
D 3.2-1 10/10/01
Gaseous Effluents Concentration D 3.2.1 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.1.1 Perform an assessment of compliance for 31 days DLCO 3.2.1(b).
Noble Gases Dose D 3.2.2 D 3.2 GASEOUS EFFLUENTS D 3.2.2 Noble Gases Dose DLCO 3.2.2 APPLICABILITY:
The air dose beyond the Site and Exclusion Area Boundary (Figure D2.a-1) due to noble gases released in gaseous effluents shall be limited to the following:
- a.
For gamma radiation, _5 5 mrad during any calender quarter and _5 10 mrad during any calender year; and
- b.
For beta radiation, _s 10 mrad during any calender quarter and
- _ 20 mrad during any calender year.
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Calculated air dose due A.1 Prepare and submit a Special 31 days following the to radioactive noble Report pursuant to end of the quarter in gases beyond the Site Specification D 5.4 to the NRC which the limit was and Exclusion Area in lieu of any other report exceeded Boundary exceeds the which identifies the cause(s)
- limit, and defines the corrective actions taken.
D 3.2-3 10/10/01
.- Noble Gases Dose D 3.2.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated air dose B.1 Prepare and submit a 31 days due to radioactive Special Report, in noble gases beyond the lieu of any other Site and Exclusion report, pursuant to Area Boundary exceeds Specification D 5.4 two times the limit, to the NRC which 1) defines actions to be taken to reduce releases and prevent recurrence and 2) results of an exposure analysis including effl uent pathways and direct radiation to determine whether the dose or dose commitment to a Member of the Public due to radiation and radioactive releases from Cooper Station during the calender year through the period covered by the calculation was
- 75 mrem to the thyroid and
- 25 mrem to the total body or any other body organ.
SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.2.1 Perform an assessment of compliance for 31 days DLCO 3.2.2.
Iodine and Particulates D 3.2.3 D 3.2 GASEOUS EFFLUENTS D 3.2.3 Iodine and Particulates DLCO 3.2.3 The dose to a Member of the Public due to 1-131, 1-133 and radioactive material in particulate form having a half-life - 8 days in gaseous effluents beyond the Site and Exclusion Area Boundary (Figure D2.a-1) shall be limited to:
- a.
- 7.5 mrem to any organ during any calender quarter; and
- b.
_ 15 mrem to any organ during any calender year.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Calculated dose due to I-A.1 Prepare and submit a Special 31 days following the 131, 1-133 and Report, in lieu of any other end of the quarter in radioactive material in report, pursuant to which the limit was particulate form having a Specification D 5.4 to the NRC exceeded.
half-life > 8 days beyond which identifies the cause(s) the Site and Exclusion for exceeding the limit(s) and Area Boundary exceeds describes the corrective action the limit, taken.
D 3.2-5 10/10/01
lodi ne-and Parti cul ates T 1.
D 3.2.3 CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose due to B.1 Prepare and submit a 31 days 1-131, 1-133 and Special Report, in radioactive material lieu of any other in particulate form report, pursuant to having a half-life Specification D 5.4
> 8 days beyond the to the NRC which 1)
Site and Exclusion defines actions to be Area Boundary exceeds taken to reduce two times the limit, releases and prevent recurrence and 2) results of an exposure analysis including effluent pathways and direct radiation to determine whether the dose or dose commitment to a Member of the Public due to radiation and radioactive releases from Cooper Station was
- 75 mrem to the thyroid and
- 25 mrem to the total body or any other body organ.
SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.3.1 Perform radioactive gaseous waste sampling In accordance and activity analysis on effluents other with than noble gases.
Table D3.2.3-1 DSR 3.2.3.2 Perform a dose assessment to determine 31 days compliance with DLCO 3.2.3.
Iodine and Particulates D 3.2.3 Table D3.2.3-1 (Page 1 of 3)
Radioactive Gaseous Waste Sampling and Analysis GASEOUS SAMPLE LOWER LIMIT RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE OF DETECTION(LLD)
TYPE TYPE FREQUENCY FREQUENCY ANALYSIS (h)
- 1. Elevated Release Point (ERP)
- 2. Reactor Building Vent
- 3.
Augmented Radwaste Building Vent Grab Sample Grab Sample Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Noble Gas Monitor Grab Sample Grab Sample Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Noble Gas Monitor Grab Sample Grab Sample Charcoal Sample 31 days (c) 31 days (c)
Principal Gamma Emitters (f) 92 days (a) 92 days H-3 Continuous (b) 7 days (d) 1-131 1-133 Continuous (b) 7 days (d)
Principal Gamma Emitters (f)
(1-131, Others)
Continuous (b) 92 days Sr-89 Sr-90 Gross Alpha Continuous (b)
Continuous Gross Noble Gases (g)
(Beta, Gamma) 31 days (c) 31 days (c)
Principal Gamma Emitters (0 92 days (a) 92 days H-3 Continuous (b) 7 days (d) 1-131 1-133 Continuous (b) 7 days (d)
Principal Gamma Emitters (f)
(1-131, Others)
Continuous (b) 92 days Sr-89 Sr-90 Gross Alpha Continuous (b)
Continuous Gross Noble Gases (g)
(Beta, Gamma) 31 days (c) 31 days (c)
Principal Gamma Emitters (f) 92 days (a) 92 days H-3 Continuous (b) 7 days (d) 1-131 I X 10- IJCi/mn (i)
I x 10- pCi/nm I 012 Jfi 1 x 10-1 PCi/mI I X 10"0 PCi/ml I X 10-1 pCi/ml (i) 1 X 10" 11Ci/mI 1 X 10-pCi/mil I X 10-6 pCi/ml I X 10' pCi/nmi (i)
I X 10-6 pCi/ml 1 X 10"12 IJCi/mi 1 X 10- 0PCi/mI 1 X 10'1 pCi/ml (i) 1 X 10"11 pCi/mI 1 X 1011,pCi/ml 1 X 10-11 pCi/ml 1 x 10o pCi/l I x 10o pCi/ml (i) 1 X 10-6 pCi/ml I X 10-12 pCi/mi wlhwujkz-REVISION 1 D 3.2-7 CNS ODAM
Iodine and Particulates D 3.2.3 Table D3.2.3-1 (Page 2 of 3)
Radioactive Gaseous Waste Sampling and Analysis GASEOUS SAMPLE LOWER RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE LIMIT OF TYPE TYPE FREQUENCY FREQUENCY ANALYSIS DETECTION (LLD)nh)
- 3.
(continued)
- 4. Turbine Building Vent (Gaseous)
- 5. Multi Purpose Facility (MPF)
Building Vent (Gaseous)
Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Noble Gas Monitor Grab Sample Grab Sample Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Noble Gas Monitor Charcoal Sample Particulate Sample Composite Particulate Sample (e)
Continuous (b)
Continuous (b)
Continuous (b) 7 days (d) 7 days (d) 92 days 1-133 Principal Gamma Emitters (f)
(1-131, Others)
Sr-89 Sr-90 Gross Alpha Continuous (b)
Continuous Gross Noble Gases (g)
(Beta, Gamma) 31 days (c) 31 days (c)
Principal Gamma Emitters (1) 92 days (a) 92 days H-3 Continuous (b) 7 days (d) 1-131 1-133 Continuous (b)
Continuous (b)
Continuous (b)
Continuous (b)
Continuous (b)
Continuous (b) 7 days (d)
Principal Gamma Emitters (f)
(1-131, Others) 92 days Sr-89 Sr-90 Gross Alpha Continuous Gross Noble Gases (g)
(Beta, Gamma) 7 days (d) 1-131 1-133 7 days (d)
Principal Gamma Emitters (f)
(1-131, Others) 92 days Sr-89 Sr-90 Gross Alpha I X 10"10 pCi/ml I X 10-1 pCi/ml (i) 1 X 10-11 pCi/ml I X 10"13 pCi/mI Sx 10-'1 pCi/ml I X 10-6 pCi/ml 1 X 10-4 liCi/ml (i)
I X 10- pCi/ml I x 10"Ic PCi/mI I X 10"]° PCi/rml I X 1011 pCi/mlI (i) 1 X 1011 PCi/mnl I X 10"-1 pCi/ml I X 10-" PCi/mI I X 10 pCi/mI I X 10-12 pCi/ml I X 10- 0 pCi/mI 1 X 10"1 pCi/mI (i) 1 X 10"1 pCi/mI I X 10-"1 pCi/ml I X 10"1 PCi/ml (a) A H-3 grab sample will also be taken when the reactor vessel head is removed. This sample will be taken at the ERP or Reactor Building Vent whichever will be representative dependent upon the head removal vacuum procedure.
(b) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications D 3.2.1, D 3.2.2 and D 3.2.3.
(continued)
Iodine and Particulates D 3.2.3 Table D3.2.3-1 (Page 3 of 3)
Radioactive Gaseous Waste Sampling and Analysis (c) Analyses shall also be performed following an increase as indicated by the gaseous release monitor of greater than 50% in the steady state release, after factoring out increases due to power changes or other operational occurrences, which could alter the mixture of radionuclides.
(d) Analysis shall also be performed following an increase as indicated by the gaseous release monitor of greater than 50% in the steady state release, after factoring out increases due to power changes or other operational occurrences, which could alter the mixture of radionuclides.
When samples collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or less are analyzed, the corresponding LLD's may be increased by a factor of 10.
(e) A quarterly composite particulate sample shall include a portion of each week's particulate samples collected during the quarter.
(f) The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides: Kr-87, Kr-88, Xe 133, Xe-133m, Xe-135, and Xe-138 for the gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144 for particulate emissions. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the LLD for the analyses should not be reported as being present at the LLD level for that nuclide. When unusual circumstances cause LLD's higher than required for more than 31 days, the reasons shall be documented in the Radioactive Effluent Release Report.
(g) The noble gas continuous monitor shall be calibrated using laboratory analysis of the grab samples from Table D3.2.3-1 or using reference sources.
(h) The LLD is the smallest concentration of radioactive material in sample that will be detected with 95% probability (5% probability of falsely concluding that a blank observation represents a "real" signal.)
For a particular measurement system (which may include radiochemical separation):
(4.66)(s b)
LLD =
(E)(V)(2.22) (Y)e-' t Where:
LLD is the "a priori" lower limit of detection as described above (as picocurie per unit mass or volume),
sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation),
V is the sample size (in units of mass or volume),
2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable),
X is the radioactive decay constant for the particular radionuclide, and t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).
The value ofsb used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples. Typical values of E, V, Y and t shall be used in the calculation.
(i) For certain radionuclides with low gamma yield or low energies, or for certain radionuclide mixtures, it may not be possible to measure radionuclides in concentrations near the LLD. Under these circumstances, the LLD may be increased inversely proportional to the magnitude of the gamma yield (.e, I x 101/I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the LLD, as calculated in this manner for a specific radionuclide, be greater than 10% of the values specified in 10 CFR 20, Appendix B, Table 2, Column 1.
Offgas Treatment System D 3.2.4 D 3.2 GASEOUS EFFLUENTS D 3.2.4 Offgas Treatment System DLCO 3.2.4 APPLICABILITY:
Gaseous releases discharged through the Offgas Treatment System shall have at least one train of charcoal adsorbers in service.
Main condenser air ejector in service, except during startup or shutdown with reactor -- 10% rated power or when system cannot function due to low offgas flow.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Gaseous releases A.1 Restore release of gaseous 7 days discharged without either discharge via charcoal train of charcoal adsorbers.
adsorbers in service.
B.
Required Action and B.1 Prepare and submit a Special 31 days following the associated Completion Report pursuant to end of the quarter in Time not met.
Specification D 5.4 to the NRC which the release which identifies the inoperable occurred equipment and describes the corrective action taken.
D 3.2-10 10110/01
Offgas* Treatment System D 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.4.1 Verify operation of the Offgas Treatment In accordance System charcoal adsorbers by using the with the DSR gaseous effluent monitoring program in frequencies of D 3.3.2, Gaseous Effluent Monitoring.
D 3.3.2.
DSR 3.2.4.2 Project the prospect of compliance with Every 31 days DLCO 3.2.5, Condition B.
when radioactive material in gaseous effluent is released without treatment.
Exhaust Ventilation Treatment Systems (EVTS)
D 3.2.5 D 3.2 GASEOUS EFFLUENTS D 3.2.5 Exhaust Ventilation Treatment Systems (EVTS)
DLCO 3.2.5 APPLICABILITY:
The Exhaust Ventilation Treatment Systems (EVTS) shall be operated to treat radioactive materials in effluent air.
When radioactive material in gaseous effluent is being released via the associated pathway.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Radioactive material in A.1 Ensure DSR 3.2.5.1 is met.
31 days gaseous effluent released without treatment.
B.
Air is discharged without B.1 Prepare and submit a Special 31 days following the treatment for > 31 days.
Report pursuant to end of the quarter in Specification D 5.4 to the which the release AND NRC which identifies the occurred inoperable equipment and The projected dose to a describes the corrective Member of the Public due action taken.
to activity in air effluent via that pathway exceeds 0.3 mrem to any body organ.
Exhaust Ventil ati on-Treatment System D 3.2.5 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.5.1 Project the prospect of compliance with Every 31 days DLCO 3.2.5.
when radioactive material in gaseous effluent is released without treatment.
Hydrogen Concentration D 3.2.6 D 3.2 GASEOUS EFFLUENTS D 3.2.6 Hydrogen Concentration DLCO 3.2.6 APPLICABILITY:
The concentration of hydrogen in the augmented offgas treatment system downstream of the recombiners shall be limited to_< 2% by volume.
During augmented offgas treatment system operation.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Concentration of A.1 Restore concentration to within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hydrogen exceeds limits, limits.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.2.6.1 Verify Hydrogen concentration in the augmented offgas 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> treatment system downstream of the recombiners is within limits.
D 3.2-14 1 Ol0/10l0
Primary Containment Venting and Purging D 3.2.7 D 3.2 GASEOUS RELEASES D 3.2.7 Primary Containment Venting and Purging DLCO 3.2.7 Venting and purging of the primary containment shall be through the Standby Gas Treatment System.
NOTES-----
- 1.
This specification does not apply to Normal Ventilation.
- 2.
This specification does not apply during startup while performing primary containment inerting in accordance with Technical Specification 3.6.3.1, Primary Containment Oxygen Concentration, following a shutdown of > 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Requirement not met.
A.1 Suspend all venting and Immediately purging of the primary containment.
D 3.2-15 10/10/01
Primary Containment Venting and Purging D 3.2.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE DSR 3.2.7.1 The primary containment shall be determined to be aligned for venting or purging through the Standby Gas Treatment System.
FREQUENCY Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to venting or purging of the primary conta i nment AND 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter during venting or purging of the primary conta i nment.
I _____________________________
Liquid Effluent Monitoring D 3.3.1 D 3.3 INSTRUMENTATION D 3.3.1 Liquid Effluent Monitoring DLCO 3.3.1 APPLICABILITY:
The liquid effluent radiation monitoring instrumentation channels shown on Table D3.3.1-1 shall be OPERABLE with:
- a.
The minimum OPERABLE channel(s)in service.
- b.
The alarm and trip setpoints set to ensure that the limits of DLCO 3.1.1 are not exceeded.
According to Table D3.3.1-1.
ACTIONS
-NOTE Separate condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
Liquid effluent radiation A.1 Suspend liquid effluent Immediately monitoring instrumentation radiation release monitored channel alarm and trip by the inoperable channel.
setpoint less conservative than required.
ORR A.2 Declare channel inoperable.
Immediately CNS ODAM E-D 3.3-1 10/10/01
Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
One or more channels B.1 Enter the Condition Immediately inoperable, referenced in Table D3.3.1-1 for the channel.
AND B.2.1 Restore inoperable 31 days channel(s) to OPERABLE status.
OR B.2.2 In lieu of any other In accordance with report, explain in the the Radioactive Radioactive Effluent Effluent Release Release Report why the Report frequency.
instrument was not repaired in a timely manner.
Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C.
As required by Required C.1 Analyze a minimum of 2 Prior to initiating a Action B.1 and referenced independent samples in release in Table D3.3.1-1.
accordance with Table D3.1.1-1.
AND C.2 NOTE--...
Determination Action and Verification Action will be performed by two separate technically qualified members of the Facility Staff.
Determine and Prior to initiating a independently verify the release release rate calculations and discharge valving.
D.
As required by Required D.1 Collect and analyze a grab 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Action B.1 and referenced sample for gross beta in Table D3.3.1-1.
radioactivity or gross gamma AND radioactivity (as applicable) at a lower limit of detection Once per
_* 10- pCi/ml.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter E.
As required by Required E.1 Estimate flow rate during 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Action B. 1 and referenced actual release.
in Table D3.3.1-1.
AND Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> thereafter CNS ODAM D 3.3-3 lO/10/01l 1
Liquid Effluent Monitoring D 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F.
Required Action and F.1 Suspend liquid effluent Immediately associated Completion releases monitored by the Time for Condition C or E inoperable channel(s).
not met.
Liquid Effluent Monitoring D 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.1.1 Perform CHANNEL CHECK.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3.1.2 Perform CHANNEL CHECK for each channel to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on any demonstrate OPERABILITY by verifying indication of day on which flow during periods of release, continuous, periodic, or batch releases are made DSR 3.3.1.3 Perform SOURCE CHECK.
Completed prior to each release DSR 3.3.1.4 Perform SOURCE CHECK.
31 days DSR 3.3.1.5 Perform CHANNEL CALIBRATION 18 months DSR 3.3.1.6 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate automatic isolation of the pathway for instrument indication levels measured above the alarm/trip setpoint and circuit failure; and control room alarm annunciation for instrument indication levels measured above the alarm/trip setpoint, circuit failure and instrument indicating downscale failure.
Liquid Effluent Monitoring D 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.1.7 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate control room alarm annunciation for instrument indication levels measured above the alarm/trip setpoint, circuit failure, instrument indicating downscale failure, and instrument controls not set in operate mode.
DSR 3.3.1.8 Perform CHANNEL FUNCTIONAL TEST.
184 days DSR 3.3.1.9 Perform LOGIC SYSTEM FUNCTIONAL TEST 184 days CNS ODAM D 3,3-6 1O/10/01 1
Liquid Effluent Monitoring D 3.3.1 Table D3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITIONS OPERABLE ACTION B.1 REQUIREMENTS
- 1. Gross Beta or Gamma Radioactivity Monitors Providing Automatic Isolation
- a. Liquid Radwaste (a) 1(b)
C DSR 3.3.1.1 Effluent Line DSR 3.3.1.3 DSR 3.3.1.5 DSR 3.3.1.6 DSR 3.3.1.9
- 2. Gross Beta or Gamma Radioactivity Monitors Providing Alarm but not Providing Automatic Isolation
- a. Service Water System (a) 1 D
DSR 3.3.1.1 Effluent Une DSR 3.3.1.4 DSR 3.3.1.5 DSR 3.3.1.7
- 3. Flow Rate Measurement Devices (a) 1 E
DSR 3.3.1.2
- a. Liquid Radwaste DSR 3.3.1.5 Effluent Une DSR 3.3.1.8 (a) During releases via this pathway.
(b) Set to alarm and automatically close the waste discharge valve prior to exceeding the limits of DLCO 3.1.1.
Gaseous Effluent Monitoring D 3.3.2 D 3.3 INSTRUMENTATION D 3.3.2 Gaseous Effluent Monitoring DLCO 3.3.2 APPLICABILITY:
The gaseous effluent radiation monitoring instrumentation channel(s) shown in Table D3.3.2-1 shall be OPERABLE with:
- a.
The minimum OPERABLE channel(s) in service.
- b.
The alarm and trip setpoints set to ensure that the limits of DLCO 3.2.1 are not exceeded.
According to Table D3.3.2-1.
ACTIONS Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
Gaseous effluent A.1 Suspend gaseous effluent Immediately radiation monitoring radiation release monitored instrumentation by inoperable channel.
channel alarm and trip setpoint less OR conservative than required.
A.2 Declare channel inoperable.
Immediately CNS ODAM 10/10/01 D 3.3-8
Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.
One or more channels B.1 Enter the Condition Immediately inoperable, referenced in Table D3.3.2-1 for the channel.
AND B.2.1 Restore inoperable 31 days channel(s) to OPERABLE status.
OR B.2.2 In lieu of any other report, In accordance with explain in the Radioactive the Radioactive Effluent Release Report Effluent Release why the instrument was not Report frequency repaired in a timely manner.
C.
As required by C.1 Ensure the offgas delay Immediately Required Action B.1 system is not bypassed.
and referenced in Table D3.3.2-1.
AND C.2 Ensure the Elevated Immediately Release Point Monitoring noble gas activity monitor is OPERABLE.
AND C.3 Restore inoperable 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> channels to OPERABLE status.
D.
Required Action and D.1 Be in MODE 2.
12 hburs associated Completion Time for Condition C not met.
D 3.3-9 10/10/01 1
Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E.
As required by E.1 Estimate flowrate.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Required Action B.1 and referenced in AND Table D3.3.2-1.
Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter F.
As required by F.1 Take grab samples.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Required Action B.1 and referenced in AND Table D3.3.2-1.
Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter AND F.2 Analyze for gross activity.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from time of sampling completion CNS ODAM 10110/011 D 3.3-10
Gaseous Effluent Monitoring D 3.3.2 CONDITION REQUIRED ACTION I COMPLETION TIME G.
As required by Required Action B.1 and referenced in Table D3.3.2-1.
Verify one Function 2.a monitor OPERABLE.
AND G.1.2 Monitor recombiner exhaust temperature.
OR G.2.1.1 Collect gas sample.
AND G.2.1.2 Analyze gas sample.
AND G.2.2.1 Verify one Function 2.a monitor OPERABLE.
OR G.2.2.2 Monitor recombiner exhaust temperature.
G.1.1 H.
Required Action and H. 1 Discontinue operation of Immediately associated Completion the augmented offgas Time for Condition G treatment system.
not met.
CNS ODAM ACTIONS Immediately Immediately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> from time of sampling completion Immediately Immediately D 3.3-11 10/10/01 1
Gaseous Effluent Monitoring D 3.3.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME As required by 1.1 Continuously collect 4 Hours Required Action B.1 samples with auxiliary and referenced in sampling equipment as Table D3.3.2-1.
required in Table D3.2.3-1.
OR 1.2.1 If auxiliary sampling Immediately equipment cannot be established within the specified completion time, enter the problem into the Corrective Action Program to evaluate particulate and iodine effluent releases.
AND 1.2.2 Report this event in the In accordance with Radioactive Effluent the Radioactive Release Report.
Effluent Release Report Frequency J.
Required Action and J.1 Discontinue effluent Immediately associated Completion releases via this pathway.
Time for Condition E, F, or I not met.
K.
Function 1.a trip K.1 Close the offgas isolation Immediately capability not valve.
maintained.
AND AND K.2 Initiate reactor shutdown.
Immediately Radiation level exceeds 1.0 ci/sec AND (prior to 30 min. delay line) for > 15 K.3 Be in MODE 4.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> consecutive minutes.
Gaseous Effluent Monitoring D 3.3.2 SURVEILLANCE REQUIREMENTS NOTES
- 1.
Refer to Table D3.3.2-1 to determine which DSRs apply for each instrument.
- 2.
When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function is maintained.
SURVEILLANCE FREQUENCY DSR 3.3.2.1 Perform CHANNEL CHECK.
24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> DSR 3.3.2.2 Perform CHANNEL CHECK.
7 days DSR 3.3.2.3 Perform SOURCE CHECK.
31 days DSR 3.3.2.4 Perform CHANNEL FUNCTIONAL TEST.
31 days DSR 3.3.2.5 Perform SOURCE CHECK.
92 days DSR 3.3.2.6 Perform CHANNEL CALIBRATION. The CHANNEL 92 days CALIBRATION shall include the use of a standard gas sample containing a percentage of hydrogen to verify accuracy of the monitoring channel in its operating range.
DSR 3.3.2.7 Perform CHANNEL FUNCTIONAL TEST.
92 days DSR 3.3.2.8 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint, circuit failure, instrument indicates a downscale failure, or instrument controls not set in operate mode.
DSR 3.3.2.9 Perform CHANNEL FUNCTIONAL TEST. The 92 days CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint or circuit failure.
Gaseous Effluent Monitoring D 3.3.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.3.2.10 Perform CHANNEL CALIBRATION. For Function 1.a, 18 months the time delay setting for closure of the steam jet air ejector isolation valves shall be < 15 minutes and trip settings shall correspond to Technical Specification 3.7.5.
DSR 3.3.2.11 Perform CHANNEL FUNCTIONAL TEST. The 18 months CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if; the instrument indicates measured levels above the alarm/trip setpoint, circuit failure, instrument indicates a downscale failure, or instrument controls not set in operate mode.
DSR 3.3.2.12 Perform LOGIC SYSTEM FUNCTIONAL TEST.
18 months CNS ODAM 10/10/011 D 3.3-14
Gaseous Effluent Monitoring D 3.3.2 Radioactive Table D3.3.2-1 (Page 1 of 3)
Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITIONS OPERABLE ACTION BA REQUIREMENTS
- a. Noble Gas Activity Monitor
- b. Effluent System Flow Rate Measuring Device
- 2. Augmented Offgas Treatment System Explosive Gas Monitoring System
- a. Hydrogen Monitor (2% monitor)
- 3. Reactor Building Ventilation Monitoring System
- a. Noble Gas Activity Monitor
- b. Iodine Sampler Cartridge
- c. Particulate Sampler Filter
- d. Effluent System Flow Rate Measuring Device
- e. Sampler Fow Rate Measuring Device
- f. Isolation Monitor (a)
I(e)
(b) 1 (c)
(b)
(b)
(b)
(b)
(b)
(d) 2 1
1 E
E (d) 1 (d)
C E
G F
DSR 3.3.2.1 DSR 3.3.2.3 DSR 3.3.2.8 DSR 3.3.2.10 DSR 3.3.2.11 DSR 3.3.2.12 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.3.2.4 DSR 3.3.2.6 DSR 3.3.2.1 DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10 DSR 3.3.2.2 DSR 3.3.2.2 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.5 DSR 3.3.2.11 (a) During operation of the steam jet air ejector (b) During releases via this pathway (c) During augmented offgas treatment system operation (d) See Technical Specification 3.3.6.2 (e) Second channel must either be OPERABLE or be in the tripped condition.
CNS ODAM D 3.3-15 (continued) 10/10/01 1
I I
Gaseous Effluent Monitoring D 3.3.2 INSTRUMENT
- 4. Elevated Release Poin Monitoring System
- a. Noble Gas Activity Monitor Table D3.3.2-1 (Page 2 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE CONDITION OPERABLE ACTION B.1 REQUIREMENT*3 It (b)
- b. Iodine Sampler (b)
Cartridge
- c. Particulate Sampler (b)
Filter
- d. Effluent System Flow (b)
Rate Measuring Device
- e. Sampler Flow Rate (b)
Measuring Device
- 5. Radwaste Building Ventilation Monitoring System
- a. Noble Gas Activity (b)
Monitor
- b. Iodine Sampler (b)
Cartridge
- c. Particulate Sampler (b)
Filter
- d. Effluent System Flow (b)
Rate Measuring Device
- e. Sampler Flow Rate (b)
Measuring Device
- 6. Turbine Building Ventilation Monitoring System
- a. Noble Gas Activity (b)
Monitor
- b. Iodine Sampler (b)
Cartridge DSR 3.3.2.1 DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10 DSR 3.3.2.2 DSR 3.3.2.2 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10 DSR 3.3.2.2 DSR 3.3.2.2 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.32.3 DSR 3.3.2.9 DSR 3.3.2.10 DSR 3.3.2.2 (b) During releases via this pathway CNS ODAM D 3.3-16 (continued)
S 10/10/01 1
Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 3 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation INSTRUMENT CONDI
- 6. (continued)
- c. Particulate Sampler (b)
Filter
- d. Effluent System Flow (b)
Rate Measuring Device
- e. Sampler Flow Rate (b)
Measuring Device
- 7. Multi Purpose Facility (MPF)
Building Ventilation Monitoring System
- a. Iodine Sampler (b)
Cartridge
- b. Particulate Sampler (b)
Filter
- c. Effluent System Flow (b)
Rate Measuring Device
- d. Sampler Flow Rate (b)
Measuring Device APPLICABILITY CONDITION OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE TION OPERABLE ACTION B.1 REQUIREMENTS E
E E
E DSR 3.3.2.2 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.2 DSR 3.3.2.2 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 DSR 3.3.2.1 DSR 3.3.2.7 DSR 3.3.2.10 (b) During releases via this pathway CNS ODAM D 3.3-17 10/10/01 1
Liquid/Gaseous Effluents Dose D 3.4.1 D 3.4 LIQUID/GASEOUS DOSE D 3.4.1 Liquid/Gaseous Effluents Dose DLCO 3.4.1 The dose or dose commitment to an actual Member of the Public due to radiation and radioactive releases from Cooper Station shall be limited to
_: 75 mrem to the thyroid and :s 25 mrem to the total body or any other body organ during a calender year.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Estimated dose or dose A.1 Verify the condition resulting in Immediately commitment due to doses exceeding these limits is radiation and radioactive corrected.
releases exceeds limits.
B.
Required Action and B.1 ---------......
NOTE associated Completion This is the Special Report Time not met.
required by D 3.1.2, D 3.1.3, or D 3.2.3 supplemented with the following.
Submit a Special Report 31 days pursuant to Specification D 5.4, including information specified in 40 CFR Part 190.11(b). This submission shall be deemed a timely request for variance in accord with provisions of 40 CFR Part 190. The variance is granted until NRC staff action on the item is complete.
D 3.4-1
LiquidlGaseous Effluents Dose D 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.4.1.1 Perform a cumulative dose calculation due to 12 months radioactive material in gaseous and liquid effluents to determine compliance with DLCO 3.4.1.
D 3.4-2
Solid Radioactive Waste D 3.5.1 D 3.5 SOLID RADIOACTIVE WASTE D 3.5.1 Solid Radioactive Waste DLCO 3.5.1 The appropriate equipment of the solid radwaste system shall be OPERABLE to process radioactive waste containing liquid and liquid destined for disposal subject to 10 CFR Part 61 to a form that meets applicable requirements of 10 CFR Part 61.56 before the waste is shipped from the site.
APPLICABILITY:
During solid radwaste processing.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Container of waste does A. 1 Suspend delivery to a carrier Immediately not comply with 10 CFR for transport.
Part 61.56.
D 3.5-1 10/10/01
Solid Radioactive Waste D 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.1.1 Sample and analyze the dewatered radioactive waste Prior to for pH.
solidification of every 10th batch of dewatered waste.
DSR 3.5.1.2 Inspect solidified or dewatered radioactive waste to Prior to capping insure that there is no free standing liquid on top of each drum or High the solid waste.
Integrity Container (HIC)
DSR 3.5.1.3 Record the following information for radioactive solid In accordance waste shipped offsite during the report period for the with Radioactive Effluent Release Report per 10 CFR 50.36a the Reporting Requirements in Technical Specification 5.6.3; a)
Container burial volume, b)
Total curie quantity (determined by measurement or estimate),
c)
Principal gamma radionuclides (determined by measurement or estimate),
d)
Type of waste, and e)
Solidification agent.
Monitoring Program Compliance D4.1 D 4.0 MONITORING PROGRAM D 4.1 Monitoring Program Compliance DLCO 4.1 The radiological environmental monitoring program shall be conducted as specified in Table D4.1-1, using analytical techniques such that the detection capabilities in Table D4.1-2 are achieved.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Radiological A.1 Prepare and submit to the May 15th following environmental monitoring NRC in the Annual the end of the year program not conducted Radiological Environmental as specified in Report the reasons for not Table D4.1-1.
conducting the program in accordance with Table D4.1-1 and the plans for preventing recurrence.
B.
Environmental sampling B.I Report in the Annual May 15th following medium is not available Radiological Environmental the end of the year from a sampling location Report the cause and location as specified in where replacement samples Table D4.1-1.
were obtained.
D 4.1 -1 10/10/01
Monitoring Program Compliance T.
D 4.1 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 4.1.1 Perform radiological environmental sampling In accordance and analysis.
with Table D4.1-1 DSR 4.1.2 Conduct a land use census to identify the 12 months location of the nearest garden that is greater than 500 square feet in area and that yields edible leafy vegetables, the location of the nearest milk animal, and the location of the nearest resident in each of the 16 meteorological sectors within 3 miles of the Station.
DSR 4.1.3 Summarize results of radiological May 15th environmental analysis in the Annual following the Radiological Environmental Report.
end of the year DSR 4.1.4 Submit results of the land use census in May 15th the Annual Radiological Environmental following the Report.
end of the year REVISION 0 CNS ODAM D 4.1-2
Monitoring Program Compliance D4.1 Table D4.1-1 (Page 1 of 2)
Radiological Environmental Monitoring Program
- a. Radioiodine and Particulate
- 2. Direct Radiation EXPOSURE PATHWAY NUMBER OF SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE STATIONS COLLECTION FREQUENCY OF ANALYSIS
- 1. Airborne At least 5 locations At least 32 locations
- 3. Waterborne
- a. River Water b.Ground Water
- c. Sediment from Shoreline At least 2 locations At least 2 locations At least I location
- 4. Ingestion
- a. Milk (nearest producer)
- b. Milk (other producers)
- c. Fish At least I location At least 2 locations At least 2 locations Continuous operation of sampler with sample collection as required by dust loading but at least once per 7 days.
Thernoluminescent Dosimeters (TLD))(o exchange and read-out at least once per 92 days.
Collect a one (1) gallon grab sample at least once per 31 days.
Collect a one (1) gallon grab sample at least once per 92 days.
Two (2) times a year, once in the spring and once in the fall.
At least once per 15 days during Peak Pasture Periodc(); at least once per 31 days at other times.
At least once per 92 days.
Two times per year (once in the summer and once in the fall).
Attempt to include the following:
Radioiodine canister: Analyze at least once per 7 days for 1-131.
Particulate sample: Analyze for gross beta radioactivity
ý 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following filter change. Perform gamma isotopca) analysis on each sample in which gross beta activity is > 10 times the yearly mean of control samples.
Perform gamma isotopic(a) analysis on composite (by location) sample at least once per 92 days.
Gamma dose: At least once per 92 days.
Gamma isotopic(a) analysis of each sample. Composite grab sample for tritium analysis at least once per 92 days.
Gamma isotopic(") and tritium analysis of each sample.
Gamma isotopic(a) analysis of each sample.
Gamma isotopic(s) and 1-131 analysis of each sample.
Gamma isotopic(a) and 1-131 analysis of each sample.
Gamma isotopic(a) analysis on edible portions.
- 1. Bottom feeding species
- 2. Middle-Top feeding species (continued)
Monitoring Program Compliance D4.1 Table D4.1-1 (Page 2 of 2)
Radiological Environmental Monitoring Program EXPOSURE PATHWAY NUMBER OF SAMPLING AND TYPE AND FREQUENCY AND/OR SAMPLE SAMPLE STATIONS COLLECTION FREQUENCY OF ANALYSIS
- 4. (continued)
- d. Food Products Samples of three different Monthly when available.
Gamma isotopic(a) and 1-131 kinds of broad leaf analysis vegetation grown nearest each of two different offsite locations of highest predicted annual average ground-level DIQ if milk sampling is not performed.
One sample of each of the Monthly when available.
Gamma isotopic() and 1-131 similar broad leaf analysis vegetation grown 15-30 km distant in the least prevalent wind direction if milk sampling is not performed.
(a) Gamma isotopic analysis refers to high resolution gamma spectrum analysis as follows: the sample is scanned for gamma-ray activity. If no activity is found for a selected nuclide, the detection sensitivity for that nuclide will be calculated using the counting time, detector efficiency, gamma energy, geometry, and detector background appropriate to the particular sample in question. The following nineteen (19) nuclides shall be analyzed routinely:
BaLa-140 Be-7 Ce-141 Ce-144 Co-58 Co-60 Cs-134 Cs-137 Fe-59 1-131 K-40 Nb-95 Mn-54 Ra-226 Ru-103 Ru-106 Th-228 Zn-65 Zr-95 Any radionuclide detected, i.e., having a measured concentration greater than the LLD, whether or not it is one of the 19 nuclides listed above, shall be regarded as present in the sample.
(b) Thermoluminescent Dosimeters (TLD) is a single phosphore. Two or more phosphores in one package are considered to be two or more dosimeters.
(c) Peak Pasture Period is June 1 through September 30 of each year.
Monitoring Program Compliance D4.1 Table D4.1-2 (Page 1 of 2)
Detection Capabilities for Environmental Sample Analysis LOWER LIMIT OF DETECTION(s) (LLD) (b)
AIRBORNE PARTICULATE FOOD WATER OR GAS FISH MILK PRODUCTS SEDIMENT ANALYSIS (pCi/I)
(pCi/m 3)
(pCi/kg, wet)
(pCi/I)
(pCi/kg, wet)
(pCi/kg, dry) gross beta 4
1x 10-.2 H-3 2000 Mn-54 15 130 Fe-59 30 260 Co-58 15 130 Co-60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 1-131 1 (c) 7 x 10' 2 1
60 Cs-134 15 5x10' 2 130 15 60 150 Cs-137 18 6x10"'
150 18 80 180 Ba-140 60 60 La-140 15 15 (a) This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
8/26/99 D 4.1-5
Monitoring Program Compliance D4.1 Table D4.1-2 (Page 2 of 2)
Detection Capabilities for Environmental Sample Analysis (b) The LLD is the "a priori" smallest concentration of radioactive material in a sample that will be detected with 95% probability (5% probability of falsely concluding that a blank observation represents a "real" signal).
For a particular measurement system (which may include radiochemical separation):
(4.66)(s b)
LLD =
(E)(V)(2.22) (Y) (e-""t)
Where:
LLD is the "a priori" lower limit of detection as described above (as picocurie per unit mass or volume),
s, is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),
E is the counting efficiency (as counts per transformation),
V is the sample size (in units of mass or volume),
2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable)
I is the radioactive decay constant for the particular radionuclide, and at is the elapsed time between sample collection (or midpoint of the sample collection period) and time of counting.
The value of %, used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g.,
potassium-40 in milk samples).
Analysis shall be performed in such a manner that the stated LLD's will be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLD's unachievable. In such cases, the contributing factors will be identified and descnried in the Annual Radiological Environmental Report.
(c)
LLD for drinking water.
Monitoring Program Concentration D 4.2 D 4.0 MONITORING PROGRAM D 4.2 Monitoring Program Concentration DLCO 4.2 Radioactivity concentrations in sampled medium from the radiological environmental monitoring program shall not exceed values specified in Table D5.4-1 when averaged over a calender quarter.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Radioactivity A. I Prepare and submit to the 31 days following the concentrations of NRC a Special Report in end of the quarter sampled medium from accordance with Specification the radiological D 5.4 which includes an environmental monitoring evaluation of any release program exceeds values conditions, environmental specified in Table D5.4-1, factors or other conditions averaged over a calender which caused the value(s) to quarter which is be exceeded.
attributable to release(s) from the Station.
B.
Radioactivity B.1 Report and explain in the May 15th following concentrations of Annual Radiological the end of the year sampled medium from Environmental Report the the radiological results of the sample(s).
environmental monitoring program exceeds values specified in Table D5.4-1, averaged over a calender quarter which is not attributable to release(s) from the Station.
Monitoring Program Dose D 4.3 D 4.0 MONITORING PROGRAM D 4.3 Monitoring Program Dose DLCO 4.3 The calculated personal dose associated with sampled exposure pathway(s) shall not exceed 120% of the calculated dose at the maximum dose location associated with like pathways at a location where sampling is conducted as specified in Table D4.1-1.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Location(s) identified at A.1 NOTE which the calculated dose Only applicable if associated with the samples are reasonably exposure pathway(s) attainable at the new exceeds 120% of the location.
calculated dose at the maximum dose location associated with like Add new sampling 92 days pathways at a location location(s) identified where sampling is having maximum conducted as specified in exposure potential to the Table D4.1-1.
radiological environmental monitoring program and Table D4.1-1.
AND A.2 Describe change made to May 15th following Table D4.1-1 in the the end of the year Annual Radiological Environmental Report.
D 4.3-1 10/10/01
Interlaboratory Comparison Program D5.1 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.1 Interlaboratory Comparison Program DLCO 5.1 Analyses shall be performed on radioactive materials supplied as part of the Interlaboratory Comparison Program.
APPLICABILITY:
At all times.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Analyses not performed.
A.1 Report to the NRC in the May 15th Annual Radiological following the end of Environmental Report the the year corrective actions taken to prevent recurrence.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 5.1.1 Submit a brief summary of the results obtained from May 15th the Interlaboratory Comparison Program in the following the end Annual Radiological Environmental Report pursuant of the year to Technical Specification 5.6.2.
Annual Radiological Environmental Report D 5.2 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.2 Annual Radiological Environmental Report The Annual Radiological Environmental Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Assessment Manual (ODAM), and in 10 CFR50, Appendix I, Section IV.B.2, IV.B.3, and IV.C.
The Annual Radiological Environmental Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODAM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in Regulatory Guide 4.8, December 1975. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.
Details of the report shall include the following:
- a.
A summary of doses to a Member of the Public beyond the Site and Exclusion Area Boundary due to Cooper Nuclear Station aqueous and airborne radioactive effluents, calculated in accordance with methods compatible with the ODAM.
- b.
A summary of the results of the land use census required in DSR 4.1.2.
- c.
Summarized and tabulated results in the format of Table D5.2-1 of analysis of samples required by the radiological environmental monitoring program, and taken during the report period.
- d.
A summary description of the radiological environmental monitoring program including any changes; a map of all sampling locations keyed to a table giving distances and directions from the reactor; and, the results of participation in the Interlaboratory Comparison Program, required by D 5.1.
- e.
Summaries, interpretations, and analysis of trends of the results of the radiological environmental monitoring program for the reporting period.
TABLE 05.2-1 ENVIRONMENTAL RADIOLOGICAL MONITORING PROGRAM
SUMMARY
Name of Facility Cooper Nuclear Station Docket No.
Location of Facility Nemaha. Nebraska (County, State)
Reporting Period Type &
Lower Limit All Indicator Control Medium of Pathways Total No.
of Locations Location with Highest Annual Mean Locations No. of Sampled of Analyses Detection(l)
Mean[](2)
Name Mean[E(2)
Mean[I(2)
Reportable (Unit of Measurement)
Performed (LLD)
Ranae (2)
Distance & Direction Range (2)
Range (2)
Occurrences tes Table Notes :
(1)
Nominal Lower Limit of Detection (LLD).
(2)
Mean and Range based upon detectable measurements only.
Fraction of detectable measurements at specified location indicated in brackets U.
I 50-298
Annual Radioactive Effluent Release Report D 5.3 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.3 Radioactive Effluent Release Report The Radioactive Release Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODAM and the Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.
The Radioactive Effluent Release Report shall be submitted to the NRC by May 1 of each year and shall include the following:
- a.
A summary by calendar quarter of the quantities of radioactive liquid and gaseous effluents released from the Station, reported in the format recommended in Regulatory Guide 1.21, Appendix B, Tables 1 and 2.
- b.
A summary of radioactive solid waste shipped from CNS, including information provided in DSR 3.5.1.3.
- c.
A summary of meteorological data collected during the year.
- d.
A list and brief description of each unplanned release of gaseous or liquid radioactive effluent that causes a limit in DLCO 3.1.1, DLCO 3.1.3, DLCO 3.2.1, DLCO 3.2.2 or DLCO 3.2.3 to be exceeded.
- e.
Calculated offsite dose to humans resulting from the release of effluents and their subsequent dispersion on the atmosphere reported in accordance with Regulatory Guide 1.21.
- f.
A summary of changes made to the CNS Process Control Program. The summary shall contain sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; a determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and documentation of the fact that the change has been reviewed and found acceptable by the SORC.
Annual Radioactive Effluent Release Report D5.3 D 5.3 Radioactive Effluent Release Report (continued)
- g.
Changes made to the CNS Offsite Dose Assessment Manual shall be submitted to the NRC in the form of a complete, legible copy of the entire ODAM as part of, or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODAM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.
Special Repor t s D5.4 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.4 Special Reports Special reports shall be submitted to the Director, Nuclear Reactor Regulation, USNRC, Washington, D.C. 20555 and to the NRC Regional Administrator within the time period specified for each report.
Special reports (in lieu of Licensee Event Reports) may be required covering inspections, test and maintenance activities.
These special reports are determined on an individual basis for each unit and their preparation and submittal are designated in the Offsite Dose Assessment Manual.
A special report is required if measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table D5.4-1 when averaged over any calendar quarter sampling period.
When more than one of the radionuclides in Table D5.4-1 are detected in the sampling medium, this report shall be submitted if:
Concentration (1) + Concentration (2) +
1.0 Limit Level (1)
Limit Level (2) 1 When radionuclides other than those in Table D5.4-1 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of DLCO 3.1.3 and 3.2.3.
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Report.
Special Reports D5.4 Table D5.4-1 Reporting Levels for Radioactivity Concentrations in Environmental Samples Reporting Levels AIRBORNE PARTICULATE OR FISH FOOD PRODUCTS ANALYSIS WATER (pCi/1)
GASES (pCi/m3)
(pCi/Kg, Wet)
MILK (pCi/1)
(pCi/Kg, Wet)
H-3 2E + 4(a) 3E + 4(c)
Mn-54 IE+3 3E + 4 Fe-59 4E + 2 1E + 4 Co-58 IE + 3 3E + 4 Co-60 3E + 2 1E+4 Zn-65 3E + 2 2E + 4 Zr-Nb-95 4E + 2(b) 1-131 2
0.9 3
IE + 2 Cs-134 30 10 1E + 3 60 1E + 3 Cs-137 50 20 2E + 3 70 2E + 3 Ba-La-140 2E + 2(b) 3E + 2(b)
For drinking water samples. This is the 40 CFR 141 value.
Concentration of parent or daughter.
For samples of water not used as a source of drinking water.
REVISION 0 (a)
(b)
(c)
Major Changes to Radioactive Waste Treatment Systems D 5.5 D 5.0 MISCELLANEOUS PROGRAMS/REPORTS D 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid)
The radioactive waste treatment systems (liquid, gaseous, and solid) are those systems described in the facility Safety Analysis Report and amendments thereto, which are used to maintain that control over radioactive materials in gaseous and liquid effluents and in solid waste packaged for offsite shipment required to meet the DLCO's set forth in Specifications D 3.1.1, D 3.1.2, D 3.1.3, D 3.1.4, D 3.2.1, D 3.2.2, D 3.2.3, D 3.2.4. D 3.2.5, D 3.2.6, D 3.2.7, D 3.3.2, D 3.4.1, and D 3.5.1.
The NRC is notified of major changes to these systems under the provisions of 10 CFR Part 50.59 and Part 50.71 (USAR revisions).
OFFSITE DOSE ASSESSMENT MANUAL APPENDIX D TABLE OF CONTENTS - BASES B 3.0 ODAM SPECIFICATION APPLICABILITY...........................
B 3.0-1 B 3.1 LIQUID EFFLUENTS B 3.1.1 Liquid Effluents Concentration..............................
B 3.1-1 B 3.1.2 Liquid Waste Concentration................................
B 3.1-2 B 3.1.3 Liquid Effluents Dose.....................................
B 3.1-5 B 3.1.4 Outside Temporary Storage of Radioactive Liquid...............
B 3.1-9 B 3.2 GASEOUS EFFLUENTS B 3.2.1 Gaseous Effluents Concentration............................
B 3.2-1 B 3.2.2 Noble Gas Dose.........................................
B 3.2-2 B 3.2.3 Iodine and Particulates....................................
B 3.2-3 B 3.2.4 Offgas Treatment System.................................
B 3.2-4 B 3.2.5 Exhaust Ventilation Treatment System.......................
B 3.2-5 B 3.2.6 Hydrogen Concentration..................................
B 3.2-6 B 3.2.7 Primary Containment Venting and Purging....................
B 3.2-7 B 3.3 INSTRUMENTATION B 3.3.1 Liquid Effluent Monitoring..................................
B 3.3-1 B 3.3.2 Gaseous Effluent Monitoring..............................
B 3.3-2 B 3.4 LIQUID/GASEOUS EFFLUENTS DOSE B 3.4.1 Liquid/Gaseous Effluents Dose.............................
B 3.4-1 B 3.5 SOLID RADIOACTIVE WASTE B 3.5.1 Solid Radioactive Waste..................................
B 3.5-1 B 4.0 MONITORING PROGRAM B 4.1 Monitoring Program Compliance...........................
B 4.1-1 B 4.2 Monitoring Program Concentration...........................
B 4.2-1 B 4.3 Monitoring Program Dose.................................
B 4.3-1 B 5.0 MISCELLANEOUS PROGRAMS/REPORTS B 5.1 Interlaboratory Comparison Program.........................
B 5.1-1 B 5.2 Annual Radiological Environmental Report....................
B 5.2-1 B 5.3 Radiological Effluent Release Report.........................
B 5.3-1 B 5.4 Special Reports.........................................
B 5.4-1 B 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid)........................
ODAM Specification Applicability B 3.0 B 3.0 ODAM Specification Applicability BASES DLCOs 3.0.1, 3.0.2, and 3.0.5, and DSRs 3.0.1, 3.0.2, and 3.0.3 reflect parallel requirements in the Technical Specifications. Refer the Technical Specification Bases for appropriate discussion.
ODAM Specification DLCO 3.0.3, in lieu of imposing a plant shutdown as paralleled in Technical Specification 3.0.3, requires: (a) an Action to initiate efforts to restore compliance with the ODAM or associated Actions; and (b) an Action that requires entering the circumstances into the Corrective Action Program (CAP). These actions ensure that the appropriate actions continue to be focused on and that the circumstances concerning failure to comply with the ODAM Actions would be reviewed. This review will be conducted in accordance with the procedural guidance for CAP Notifications.
There are no ODAM 3.0 Specifications that parallel Technical Specification LCO 3.0.4 or SR 3.0.4. Restrictions in entering MODES or other specified conditions in the Applicability have historically not been applied to ODAM Specifications. There are also no ODAM 3.0 Specifications that parallel Technical Specification LCO 3.0.6 and LCO 3.0.7, which allow for exceptions and revisions of other Technical Specifications. They are not applicable to the ODAM since it is not permitted to allow the ODAM to revise a Technical Specification.
(Note that there currently are no identified ODAM DLCOs that support Technical Specification systems; however, this discussion is presented to address the philosophy that would be applied.) An allowance similar to Technical Specification LCO 3.0.6 does not apply to the ODAM. When a Technical Specification supported system LCO is discovered to be not met solely due to a ODAM support system DLCO not met, appropriate Technical Specification ACTIONS are required to be entered immediately. This applies even in instances where the ODAM contains a delay prior to declaring a Technical Specification supported system inoperable. In this case, certain ODAM inoperabilities may not directly impact the OPERABILITY of the Technical Specification supported system and delayed declaration of inoperability of the supported system is acceptable. In other cases, discovered support system inoperabilities that directly result in supported system inability to perform the safety function, should result in immediate declaration of inoperability of the supported system.
Technical Specification LCO 3.0.7 has no parallel in the ODAM since it provides for explicit changes to specified Technical Specifications by the Section 3.10 Specifications. However, in the event that LCO 3.0.7 provides for changes to the Technical Specification MODE definitions by the Section 3.10 Specifications, the revised MODE definitions apply to all plant references, including ODAM references.
Liquid Effluents Concentration B 3.1.1 B 3.1 LIQUID EFFLUENTS B 3.1.1 Liquid Effluents Concentration BASES This specification is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than the concentration levels specified in 10 CFR Part 20.1302. This limitation provides additional assurance that the levels of radioactive materials in bodies of water outside the site will result in exposures within (1) the Section IV.A guides on technical specifications in Appendix I, 10 CFR Part 50, for an individual and (2) the limits of 10 CFR Part 20.1301 and 20.1302(b)(2)(i) to the population. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2.
Since Service Water is not a normal or expected source of significant radioactive release, routine sampling and monitoring for radioactivity is precautionary.
An activity concentration of 3 x 10-6 pCi/mlin Service Water effluent is diluted in the discharge canal to about 1.5% of the 10 CFR 20 Appendix B Table 2 Column 2 concentration with only one circulating water pump operating. During normal Station operation the dilution would be even greater. By monitoring Service Water effluent continuously for radioactivity and by confirmatory sampling weekly, reasonable assurance that its activity concentration can be kept to a small fraction of the 10 CFR Part 20.1302 limit and within the Specification D 3.1.3 limit is provided.
By monitoring Service Water continuously and liquid radwaste continuously during discharge with the monitor set to alarm or trip before the limit specified in 10 CFR 20.1302 is exceeded, reasonable assurance of compliance with Specification D 3.1.1 is provided. Verification that radioactivity in liquid effluent averaged only a small fraction of the concentration limit is provided by calculations demonstrating compliance with Specification D 3.1.3.
Compliance with 10 CFR Part 20.1302(b)(2)(i) implies that the concentration limit represented by 10 CFR Part 20, Appendix B, Table 2 will be met within a suitable and reasonable averaging time for assessing compliance. That averaging time is dependent upon the resolving time of the measurements or estimates which are used to evaluate compliance. Assessment of compliance is done by sampling and analysis according to DSR 3.1.1.2, by estimating or measuring the maximum release flow and the minimum dilution flow coincident during the period of release represented by the sample, and by computing the concentration as a fraction of the limit beyond the site and exclusion area boundary periodically on the basis of these data.
Reporting by Special Reports and other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Liquid Waste Concentration B 3.1.2 B 3.1 LIQUID EFFLUENTS B 3.1.2 Liquid Waste Concentration BASES Specification D 3.1.2 implements the requirements of 10 CFR Part 50.36a(a)(1) that operating procedures be established and followed and that equipment be maintained and used to keep releases to the environment as low as is reasonably achievable.
The OPERABILITY of the liquid radwaste treatment system ensures that the appropriate portions will be available for use whenever liquid effluents require treatment prior to release to the environment.
The specification that the portions of the system which were used to establish compliance with the design objectives in 10 CFR Part 50, Appendix I, Section II be used when specified provides reasonable assurance that releases of radioactive material in liquid effluent will be kept as low as is reasonably achievable.
The activity concentration, 0.01 pCi/ml, below which liquid radwaste treatment would not be cost beneficial, and therefore not required, is demonstrated below:
The quantity of radioactive material in liquid effluent released annually from Cooper Station has been calculated to be' total iodines 3.65 curies total others (less H3) 0.7 total 4.35 curies The population dose commitment resulting from the radioactive material in liquid effluent released annually has been calculated to be thyroid 1.95 manrem total body 0.56 total 2.5 manrem Therefore, population doses are about 0.5 manrem per curie of iodine released and about 0.8 manrem per curie of other radionuclides (less WI) released in liquids.
It would be conservative to assume one manrem committed per curie released in liquid effluent.
The volume of liquid waste processed and intended for discharge is estimated to be:
Low Purity Waste 5700 gal/day 1.8 x 106 gal/yr Chem Waste + Demin Regenerant Waste 4000 gal/day 1.2 x 106 gal/yr (continued)
Liquid Waste Concentration B 3.1.2 BASES (continued)
The annual credit for to be:
costs to operate the radwaste processing equipment, neglecting capital recovery, are estimated according to Regulatory Guide 1.110 Dirty Waste lonex Evaporator
$ 88,000/yr
$114,000/yr Unit volume operating costs are about:
Cost to ion exchanger =
$ 88.000 1.8E+6 gal Cost to evaporate
$1.26 g00
- 1.
2E+6 gal
= $0.05/gal
= $0.10/gal Assuming the cost-benefit balance is $1,000 expenditure per manrem reduction and assuming treatment removes all radioactivity from the liquid, then (1) the activity concentration in a batch below which treatment cost-beneficial is
$88,000 1.8E+6 gal x 3785 mL gaI 1 curie 106 pCi x
x x
manrem curie C = 0.013 pCi/ml (2) the activity concentration below which evaporation is not costbeneficial is C $114,000 1 curie 106 PCi C=x x
x 1.2E+6 gal x 3785 ml-.
manrem curie gaT 1 manrem
$1,000 C = 0.025 pCi/ml Therefore, to one significant digit, radwaste treatment of liquids containing less than 0.01 pCi/ml is not justified.
(continued)
REVISION 0 is not 1 manrem
Liquid Waste Concentration B 3.1.2 BASES (continued) 1Demonstration of Compliance with 10 CFR 50 Appendix I, Revision 1 and Supplement 2, Nebraska Public Power District, Cooper Nuclear Station, January 9, 1978.
Liquid Effluents Dose B 3.1.3 B 3.1 LIQUID EFFLUENTS B 3.1.3 Liquid Effluents Dose BASES Note: The Bases discussion refers to "technical specifications" and quotes the Staff's use of "technical specifications."
The statements and opinions pre date Generic Letter 89-01, Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the relocation of procedural details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program.
Generic Letter 89-01 provides the guidance and justification for relocation of these "technical specifications" to the Offsite Dose Assessment (ODAM)
Manual and the Process Control Program (PCP).
Therefore, "technical specifications" as used in this Bases refers to ODAM Specifications.
Specifications D 3.1.3, D 3.2.2 and D 3.2.3 implement the requirements of 10 CFR Part 50.36a and of 10 CFR Part 50, Appendix I, Section IV. These specifications state ODAM LIMITING CONDITIONS FOR OPERATION (DLCO) to keep levels of radioactive materials in LWR effluents as low as is reasonably achievable.
Compliance with these specifications will also keep average releases of radioactive material in effluents at small percentages of the limits specified in 10 CFR Part 20.1301.
Surveillance Requirements provide for the measurement of releases and calculation of doses to verify compliance with the Specifications.
Action statements in these Specifications implement the requirements of 10 CFR Part 50.36(c)(2) and 10 CFR Part 50, Appendix I, Section IV.A in the event an LCO is not met.
Annual dose limitations stated in Specifications D 3.1.3, D 3.2.2 and D 3.2.3 are not strict limits as used elsewhere in the Technical Specifications (are not an immediate safety concern) but do obligate NPPD to take the applicable Required Action in Specifications D 3.1.3, D 3.2.2 and D 3.2.3.
(continued)
Liquid Effluents Dose B 3.1.3 BASES (continued) 10 CFR Part 50 contains two distinctly separate statements of requirements pertaining to effluents from nuclear power reactors.
The first concerns a description of equipment to maintain control over radioactive materials in effluents, determination of design objectives, and means to be employed to keep radioactivity in effluents ALARA.
This requirement is stated in Part 50, Section 34a and Appendix I,Section II.
Appendix I,Section III stipulates that conformance with the guidance on design objectives be demonstrated by calculations (since demonstration is expected to be prospective).
The other is a requirement for developing limiting conditions for operation in technical specifications.
It is stated in 10 CFR Part 50, Section 36a and Appendix I,Section IV. Both the intent of the Commission and the requirement are clearly stated in the Opinion of the Commission; I relevant paragraphs from that document follow:
Section 50.36a(b) of 10 CFR Part 50 provides that licensees shall be guided by certain considerations in establishing and implementing operating procedures specified in technical specifications which take into account the need for operating flexibility and at the same time ensure that the licensee will exert his best efforts to keep levels of radioactive materials in effluents as low as practicable.
The Appendix I that we adopt provides more specific guidance to licensees in this respect.
A. The Rule Section IV of Appendix I specifies action levels for the licensee.
If, for any individual light water cooled nuclear power reactor, the quantity of radioactive material actually released in effluents to unrestricted areas during any calendar quarter is such as to cause radiation exposure, calculated on the same basis as the design objective exposure, which would exceed one-half the annual design objective exposure, the licensee shall make an investigation to identify the causes of these high release rates, define and initiate a program of action to correct the situation, and report these actions to the Commission within 30 days of the end of the calendar quarter.
The conclusion of the NRC Staff in the Appendix I Rulemaking Hearing 2 agrees with that of the Commission.
The Staff recommended, "...that the limiting conditions for operation described in Appendix I,Section IV be applicable upon publication to technical specifications included in any license authorizing operation of a light water cooled nuclear power reactor..."
(p. 73).
(continued)
Liquid Effluents Dose B 3.1.3 BASES (continued)
The action to be taken by a licensee in the event a limiting condition is exceeded, is stated in Appendix I,Section IV.A and in the Opinion of the Commission.
ODAM Specifications D 3.1.3, D 3.2.2, D 3.2.3 and Surveillances DSR 3.1.3.1, 3.1.3.2, 3.2.2.1, 3.2.3.1 and 3.2.3.2 for Cooper Station conform to this requirement.
Guidance for developing limiting conditions for operation for surveillance and monitoring is included in Appendix I,Section IV.B.
Although "it is expected that the annual releases of radioactive material in effluents from light water cooled nuclear power reactors can generally be maintained within the levels set forth as numerical guides for design objectives in Section II" (Appendix I,Section IV), no recommendation was made by either the Staff in its Concluding Statement 4 or by the Commission in its Opinion 5 that design objective values should appear as technical specification limits.
The Opinion of the Commission and the statement of Appendix I are clear.
Limiting conditions of operation (LCO) related to the quantity of radioactive material in effluents released to an unrestricted area stated in technical specifications shall conform to Appendix I,Section IV.A.
Licensee action in the event an LCO is exceeded should be in accord with Section IV.A.
Finally, surveillance and monitoring of effluents and the environment should conform to Section IV.B.
With the implementation of Specification D 3.1.3 and Surveillances DSR 3.1.3.1 and 3.1.3.2, there is reasonable assurance that Station operation will not cause a radionuclide concentration in public drinking water taken from the River that exceeds the standard for anthropogenic radioactivity in community drinking water.
1NRC Commissioners, "Opinion of the Commission," in the Appendix I Rulemaking hearing, Docket Rm 502, p. 101-102, April 30, 1975.
2NRC Staff, "Concluding Statement of the Regulatory Staff," in the Appendix I Rule-making Hearing, Docket RM 502, pp. 17, 69, 73, 115, February, 1974.
3NRC Commissioners, p. 101 4NRC Staff, op. cit.
5NRC Commissioners, op. cit.
(continued)
Liquid Effluents Dose B 3.1.3 BASES (continued) 6Generic Letter 89-01, Implementation of Programmatic Controls for Radiological Effluent Technical Specifications in the Administrative Controls Section of the Technical Specifications and the relocation of procedural details of RETS to the Offsite Dose Calculation Manual or to the Process Control Program.
Outside Temporary Storage of Radioactive Liquid B 3.1.4 B 3.1 LIQUID EFFLUENTS B 3.1.4 Outside Temporary Storage of Radioactive Liquid BASES Custom Technical Specifications Bases did not exist.
Gaseous Effluents Concentration B 3.2.1 B 3.2 GASEOUS EFFLUENTS B 3.2.1 Gaseous Effluents Concentration BASES DLCO 3.2.1(a) is included to assure that a measure of control is provided over the concentration of radionuclides in air leaving the exclusion area.
Radioactive noble gases are monitored by instruments that provide a measure of release rate and cause automatic alarm when the noble gas concentration beyond the Site and Exclusion Area Boundary is expected to exceed the dose rate specified in DLCO 3.2.1(a).
With prompt action to reduce the radioactive noble gas concentration in effluent following alarm initiation, it can be maintained at a small fraction of the annual limit.
The specified release rate limits restrict the corresponding gamma and beta dose rates above background to an individual at or beyond the exclusion area boundary to
- 500 mrem/year to the total body or to
- 3000 mrem/year to the skin.
Radioiodines and radionuclides in particulate form are sampled with integrating samplers that permit assessment of the average release rate during each sample collection period.
By complying with DLCO 3.2.2 and 3.2.3 the average concentration beyond the Site and.Exclusion Area Boundary will be maintained at a small fraction of the 10 CFR Part 20.1302(b)(2)(i) concentration limit.
Noble Gases Dose B 3.2.2 B 3.2 GASEOUS EFFLUENTS B 3.2.2 Noble Gases Dose BASES Assessments of dose required by Surveillances DSR 3.2.2.1 and DSR 3.2.3.2 to verify compliance with Appendix I,Section IV is based on measured radioactivity in gaseous effluent and on calculational methods stated in the ODAM. Pathways of exposure and location of individuals are selected such that the dose to a nearby resident is unlikely to be underestimated. Dose assessment methodology described in the ODAM for gaseous effluent will be consistent with the methodology in Regulatory Guides 1.109 and 1.111. Cumulative and projected assessments of dose made during a quarter are based on historical average, or reference (the same period of record used in the design objective Appendix I evaluation) atmospheric conditions. Assessments made for the Annual Radiological Environmental Report will be based on quarterly and annual averages of atmospheric conditions during the period of release.
The bases for Specification D 3.2.2 and Surveillance DSR 3.2.2.1 are also discussed in the bases for Specification D 3.1.3 and Surveillances DSR 3.1.3.1 and 3.1.3.2.
Iodine and Particulates B 3.2.3 B 3.2 GASEOUS EFFLUENTS B 3.2.3 Iodine and Particulates BASES This bases for Specification D 3.2.3 and Surveillances DSR 3.2.3.1 and 3.2.3.2 are discussed in the bases for Specification D 3.1.3 and Surveillances DSR 3.1.3.1 and 3.1.3.2.
Offgas Treatment System B 3.2.4 B 3.2 GASEOUS EFFLUENT B 3.2.4 Offgas Treatment System BASES The OPERABILITY of the gaseous radwaste treatment system ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment.
The requirement that the appropriate portions of this system be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."
This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section lID of Appendix I to 10 CFR Part 50.
The specified limits governing the use of appropriate portions of this system are specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
Exhaust Ventilation Treatment Systems B 3.2.5 B 3.2 GASEOUS EFFLUENTS B 3.2.5 Exhaust Ventilation Treatment Systems BASES An Exhaust Ventilation Treatment System (EVTS) is a system intended to remove radioiodine or radioactive material in particulate form from gaseous effluent by passing exhaust ventilation air through charcoal absorbers and/or HEPA filters before exhausting the air to the environment.
An EVTS is not intended to affect noble gas in gaseous effluent.
Engineered Safety Feature (ESF) gaseous treatment systems are not considered to be EVTS.
The Standby Gas Treatment System is an ESF and not an EVTS.
EVTS are specifically identified in ODAM Figure 3-1.
The OPERABILITY of the exhaust ventilation treatment systems ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment.
The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."
This specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section lID of Appendix I to 10 CFR Part 50.
The specified limits governing the use of appropriate portions of the system are specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
Hydrogen Concentration B 3.2.6 B 3.2 GASEOUS EFFLUENTS B 3.2.6 Hydrogen Concentration BASES This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen and oxygen.
While the Augmented Treatment System is in service the hydrogen and oxygen concentrations are prevented from reaching the flammability limits.
Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR 50.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Containment B 3.2.7 B 3.2 GASEOUS RELEASES B 3.2.7 Primary Containment Venting and Purging BASES This specification provides reasonable assurance that releases of Iodine from drywell purging during power operations, and during startup while performing primary containment inerting within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown will not be excessively large, particularly due to Iodine spiking.
The exemptions to using the SBGT system are intended to minimize the time the SBGT system is on line while coolant temperature is greater than 200'F, hence to decrease the probability of damage to the SBGT filters that could occur from overpressurization due to a LOCA and the main purge and vent valves open.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Liquid Effluent Monitoring B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Liquid Effluent Monitoring BASES The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the release of radioactive material in liquid effluents.
The OPERABILITY and use of these instruments implements the requirements of 10 CFR Part 50, Appendix A, General Design Criteria 60, 63, and 64.
The alarm and/or trip setpoints for these instruments are calculated in the manner described in the ODAM to assure that the alarm and/or trip will occur before the limit specified in 10 CFR Part 20.1302 is exceeded.
Control of the normal liquid discharge pathway is assured by station procedures governing locked discharge valves and valve line-up verification.
The liquid radwaste monitor assures that all liquid discharged to the discharge canal does not exceed the limits of Specification D 3.1.1.
Upon sensing a high discharge level, an isolation signal is generated which closes the radwaste discharge valve.
The set point is adjustable to compensate for variable isotopic discharges and dilution flow rates.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Gaseous Effluent Monitoring B 3.3.2 B 3.3 INSTRUMENTATION B 3.3.2 Gaseous Effluent Monitoring BASES The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents.
The location of this instrumentation is indicated by a Figure in the ODAM, a
simplified flow diagram showing gaseous effluent treatment and monitoring equipment.
The alarm/trip setpoints for these instruments shall be calculated in accordance with methods in the ODAM, which have been reviewed by NRC, to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part
- 20.
The process monitoring instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the augmented offgas treatment system. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
Two air ejector off-gas monitors are provided and when their trip point is reached, cause an isolation of the air ejector off-gas line.
Isolation is initiated when both instruments reach their high trip point or one has an upscale trip and the other a downscale trip. There is a fifteen minute delay accounted for by the 30-minute holdup time of the off-gas before it reaches the stack.
Both instruments are required for trip but the instruments are so designed that any instrument failure gives a downscale trip.
The trip setting of 1.0 ci/sec (prior to 30 min. delay) provides an improved capability to detect fuel pin cladding failures to allow prevention of serious degradation of fuel pin cladding integrity which might result from plant operation with a misoriented or misloaded fuel assembly.
This limit is more restrictive than 0.39 ci/sec noble gas release rate at the air ejectors (after 30 min. delay) which was used as the source term for an accident analysis of the augmented off-gas system.
Using the.39 ci/sec source term, the maximum off-site total body dose would be less than the.5 rem limit.
In the event no flow rate measurement device is operable on a gaseous stream, alternative 24-hour estimates are adequate since the system design is constant flow and loss of flow is alarmed in the control room.
For this specification, reporting by Special Reports and the other reports required by the ODAM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Liquid/Gaseous Effluents Dose B 3.4.1 B 3.4 LIQUID/GASEOUS DOSE B 3.4.1 Liquid/Gaseous Effluents Dose BASES This specification is provided to meet the reporting requirements of 40 CFR Part 190. In the event an analysis is required to determine compliance with 40 CFR 190, the dose to a member of the public due to radiation direct from the station will be estimated with the aid of environmental TLD, PIC, or similar environmental radiation dosimetry.
A contribution from another fuel cycle facility is not added since there is no licensed fuel cycle facility within 50 miles of Cooper Station.
Solid Radioactive Waste B 3.5.1 B 3.5 SOLID RADIOACTIVE WASTE B 3.5.1 Solid Radioactive Waste BASES The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require materials processing and packaging prior to being shipped offsite.
This specification implements the requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A to 10 CFR Part 50.
Monitoring Program Compliance B 4.1 B 4.0 MONITORING PROGRAM B 4.1 Monitoring Program Compliance BASES The radiological environmental monitoring program, including the land use census, is conducted to satisfy the requirements of 10 CFR Part 50, Appendix I, Section IV.B.2 and 3.
The radiological monitoring program required by this specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation.
This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measureable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.
The environmental monitoring program described in Table D4.1-1 is the minimum program which will be maintained.
The Offsite Dose Assessment Manual(ODAM) describes in detail the actual monitoring program which is performed to ensure compliance with the specified minimum program.
The land use census is conducted annually to identify changes in use of the unrestricted area in order to recommend modifications in monitoring programs for evaluating individual doses from principal exposure pathways.
The need to adjust the program to current conditions and to assure that the integrity of the program is maintained are thereby provided.
Restricting the census to gardens of greater than 500 square feet provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/year) of leafy vegetables assumed in Regulatory Guide 1.109 or consumption by a child.
To determine this minimum garden size, the following assumptions were used, 1) that 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/square meter.
Monitoring Program Concentration B 4.2 B 4.0 MONITORING PROGRAM B 4.2 Monitoring Program Concentration BASES Custom Technical Specifications Bases did not exist.
Monitoring Program Dose B 4.3 B 4.0 MONITORING PROGRAM B 4.3 Monitoring Program Dose BASES Like pathways monitored (sampled) at a location, excluding the control station location(s), having the lowest associated calculated personnal dose may be deleted from Table D4.1-1 at the time the new pathway(s) and locations are added.
Interlaboratory Comparison Program B 5.1 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.1 Interlaboratory Comparison Program BASES The requirement for participation in a Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.
Annual Radiological Environmental Report B 5.2 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.2 Annual Radiological Environmental Report BASES Custom Technical Specifications Bases did not exist.
Annual Radiological Effluent Release Report B 5.3 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.3 Annual Radiological Effluent Release Report BASES Custom Technical Specifications Bases did not exist.
Special Reports RB5.4 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.4 Special Reports BASES Custom Technical Specifications Bases did not exist.
Major Changes to Radioactive Waste Treatment Systems B 5.5 B 5.0 MISCELLANOUS PROGRAMS/REPORTS B 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid)
BASES Custom Technical Specifications Bases did not exist.
REVISION 1
". " ODAM B 5.5-1
PROCESS CONTROL PROGRAM FOR COOPER NUCLEAR STATION
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A.
Purpose:
The Cooper Nuclear Station (CNS) Process Control Program (PCP) establishes the processing conditions for assuring the SOLIDIFICATION, dewatering or stabilization of CNS radioactive waste streams produced from the CNS liquid radioactive waste treatment system and from activities producing radioactive waste requiring SOLIDIFICATION, dewatering or stabilization such as decontamination system resins, irradiated components and highly contaminated equipment.
The CNS PCP is comprised of the Dewatering Process Control Program (DPCP),
SOLIDIFICATION Process Control Program (SPCP) and vendor Process Control Programs. The DPCP utilizes the CNS dewatering system to process solid wet waste streams from the CNS liquid radioactive waste treatment system or from chemical decontamination resins. The SPCP utilizes the cement solidification system to process solid wet waste from the liquid radioactive waste treatment system. Vendor Process Control Programs utilize NRC approved PCP's, stabilization processes and High Integrity Containers to process various forms of solid and liquid radioactive wastes at CNS.
B.
Definitions:
- 1.
SOLIDIFICATION, SOLIDIFY, SOLIDIFIED -The conversion of radioactive wastes from liquid systems to a solid which is as uniformly distributed as reasonably achievable with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).
- 2.
Stability - Structural stability which ensures that the waste does not degrade and (a) promote slumping, collapse, or other failure of the cap or cover over a near-surface disposal facility and thereby lead to water infiltration, or (b) impart a substantial increase in surface area of the waste form that could lead to an increase in leach rate. Stability is also a factor in limiting exposure to an inadvertent intruder since it provides greater assurance that the waste form will be recognizable and nondispersable during its hazardous lifetime. Structural stability of a waste form can be provided by the waste form itself (as with activated stainless steel components),
by processing the waste to a stable form (e.g., stabilization), or by placing the waste in a container or structure that provides stability (e.g., high integrity container or engineered structure).
C.
Discussion:
The regulation, "Licensing Requirements for Land Disposal of Radioactive Waste,"
10 CFR Part 61, establishes a waste classification system based on the radionuclide concentrations in the waste. All Class B and C wastes are required to be stabilized. Class "A wastes have lower concentrations and may be segregated without stabilization. All Class "A liquid wastes, however, require solidification or absorption to meet the free liquid requirements. Class A wastes may also be stabilized and disposed of with stabilized Class B and C wastes.
D.
References:
- 1.
CNS Operating Procedure 2.5.4.1, Solid Wet Waste Packaging, Storage and Transfer System.
- 2.
CNS Operating Procedure 2.5.4.2, Solid Wet Waste Drum Filling.
- 3.
CNS Operating Procedure 2.5.4.3, Radwaste Drum Mixer Operation.
- 4.
CNS Operating Procedure 2.5.4.4, NuPac Dewatering System.
- 5.
CNS Operating Procedure 2.5.1.3, Radwaste Low Conductivity Liquid Waste Spent Resin Tank Fluid Transfer.
- 6.
CNS Operating Procedure 2.5.2.1, Radwaste High Conductivity Liquid Waste and Waste Sludge Tank Fluid Transfer.
- 7.
CNS Operating Procedure 2.5.1.2, Radwaste Low Conductivity Liquid Waste Condensate Phase Separator Fluid Transfer.
- 8.
CNS Operating Procedure 2.5.1.7, RWCU Phase Separator Tank Transfer.
- 9.
CNS Radiological Protection Procedure 9.RW.1, Radioactive Shipments.
- 10.
CNS Radiological Protection Procedure 9.RW.2, Condensate Waste Resins, Spent Resins, RWCU Resins and Waste Sludge Classification and Listing.
- 11.
CNS Radiological Protection Procedure 9.RW.7, Waste Stream Sampling.
- 12.
Cooper Nuclear Station Offsite Dose Assessment Manual, D.3.5.1 - Solid Radioactive Waste.
- 13.
Cooper Nuclear Station Technical Specifications,5.6.3 - Radioactive Effluent Release Report.
- 14.
Cooper Nuclear Station Updated Safety Analysis Report, XIII-13.0- Process Control Program.
- 15.
10CFR61 Licensing Requirements for Land Disposal of Radioactive Waste.
- 16.
Waste Form Technical Position, Revision 1, January 1991, U. S. NRC.
- 17.
Certificate of Compliance WN-EB-01, State of Washington Department of Health.
- 18.
Certificate of Compliance WN-EB-02, State of Washington Department of Health.
- 19.
Safety Evaluation Report TP-02-P, Revision 1, U. S. NRC.
- 20.
Cooper Nuclear Station Offsite Dose Assessment Manual, D.5.3. - Radioactive Effluent Release Report.
Process Descriptions:
- 1.
SOLIDIFICATION Process Control Program
- a.
Overview The Cooper Nuclear Station (CNS) SOLIDIFICATION Process Control Program (SPCP) is a program of sampling, processing, analysis and formulation determination by which SOLIDIFICATION of radioactive waste from liquid systems is assured to be consistent with the CNS Offsite Dose Assessment Manual. Compliance with the SPCP ensures that the resultant waste form characteristics are acceptable for burial at a licensed low level radioactive waste burial facility.
The SPCP is not intended to be a substitute for the CNS Operations Manual. The SPCP outlines the general methods of sampling, processing, analysis and waste formulation during the SOLIDIFICATION of solid wet radioactive waste. The CNS Operations Manual details the actual methods by which solid wet waste is SOLIDIFIED.
Solid wet waste streams produced from the liquid radioactive waste treatment system can be SOLIDIFIED utilizing the CNS SPCP. These wet waste streams are comprised of condensate resins, spent resins, reactor water cleanup resins and waste sludges.
When regulations require SOLIDIFICATION, thesewaste streams are SOLIDIFIED in accordance with SPCP requirements. The waste processing is performed with equipment installed at CNS under the direct supervision of CNS personnel utilizing CNS approved procedures. ALARA considerations and radiological controls are incorporated into these procedures. After the waste has been processed in its burial container, the waste classification and type is determined so that the waste may be shipped to a licensed burial site in accordance with appropriate NRC, DOT, state and burial site regulations.
- b.
Operation - General:
The solid wet radioactive waste SOLIDIFICATION process is operated on a batch basis. A batch consists of all the resulting drums sequentially processed from the contents of a single source such asa phase separator.
The batch is completed when the waste source is empty or the batch process is terminated due to radioactive waste system operational constraints. Solid wet waste sources which are typically processed on a batch basis for SOLIDIFICATION include the condensate phase separators, reactor water clean-up phase separators, waste sludge tank or spent resin tank.
Waste steams being SOLIDIFIED from the condensate or reactor water cleanup phase separators, waste sludge tank, or spent resin tank are routed through centrifuge units for the purpose of dewatering the waste slurry. The wastes enter a storage hopper after dewatering in the centrifuge units. The waste in the storage hopper ranges from a relatively dry granular consistency to a wet, putty-like consistency, depending on the waste material (e.g., sludge, resin). Department of Transportation (DOT) 17H specification 55 gallon drums containing dry Portland Type I cement are transferred under the hopper and filled with the dewatered waste and demineralized water. The drum then progresses to the drum mixing section where an in-drum mixes the cement and waste.
During the mixing process additional water is added to the mixture in quantities sufficient to ensure SOLIDIFICATION.
Due to the varying degrees of moisture content subsequent to the centrifuge dewatering process, a visual inspection of the initial drums being mixed is necessary to establish the exact amount of water required to SOLIDIFY the waste batch.
Once the correct volume of water for the batch is determined the mixing system programmer is adjusted to add the correct amount of mixing water automatically. The water, cement and waste are thoroughly mixed by the mixing system's programmed sequence. After the mixing sequence is CNS PCP 8/7/00 1
complete the radiation levels are measured on each drum. The drum is then transferred to the drum storage line.
The filled drums are allowed to SOLIDIFY a minimum of twenty-four hours before being transferred from the drum storage line to the drum capping station. Here the drums are visually inspected for freestanding liquid and SOLIDIFICATION. Drums having no visible freestanding liquid and meeting SOLIDIFICATION acceptance criteria are capped, decontaminated and returned to the drum storage line in preparation for shipment to a licensed burial facility.
Drums not complying with freestanding liquid of SOLIDIFICATION requirements are retained for corrective disposition prior to capping and storage.
- c.
Operation - Specific:
1.8 +/- 0.3 cubic feet of dry Portland Type I cement is added to each DOT 17H specification 55 gallon drum. It has been demonstrated that this volume of cement will properly SOLIDIFY the waste streams processed by the drum mixing system.
The waste to be SOLIDIFIED is added in quantities sufficient to fill the drum along with approximately seven gallons of demineralized water. When the drum filling is complete the drum is transferred to the drum mixing station. A representative waste sample from one drum of each batch is taken and analyzed for pH. Experience has conclusively shown that the SOLIDIFICATION process is unaffected by pH as long as the pH of the waste remains within the range of 2 to 13.
The filled drum is mixed by an automatic, sequence controlled, in-drum mixer. During the mixing sequence approximately five gallons of additional demineralized water are added to the filled drum. When the water addition is complete and the final mixing steps are sequencing to completion, the mixer motor amps are checked to ensure sufficient water has been added.
Slight changes in the amount of water being added during the mixing sequence may be made to adjust the mixer motor amps to approximately CNS PCP 817/00 1
seven amps. These adjustments are made on the first several drums being processed from the waste batch, and are then programmed into the drum mixing sequence for the remaining drums of the batch. Upon completion of the mixing process, and prior to the cement setting, one drum from the batch is sampled in order to determine its isotopic content and distribution.
The drum's radiation level is also measured. By comparing this drum's isotopic distribution, concentration, and radiation level with the radiation readings on each of the other drums in the batch the total concentration of the radionuclides present in each drum can be determined. This data comparison method may also be used to determine waste carry-over from previous batches into the batch currently being processed.
Following a minimum of twenty-four hours after mixing, each SOLIDIFIED drum is visually inspected for freestanding liquid and SOLIDIFICATION.
Every tenth drum of the batch is quality control checked for resistance to penetration to verify SOLIDIFICATION. The penetration test verifies that the SOLIDIFIED waste has a minimum compressive strength of 50 psi. Filled drums meeting the freestanding liquid and SOLIDIFICATION criteria are capped and transferred to the drum storage line for later shipment to a licensed burial site. The packaging, classification and shipping of wastes processed via the CNS SPCP are in accordance with the applicable sections of 10CFR61, 10CFR71 and 49CFR Parts 171 through 178.
- d.
Qualification:
Qualification of the SPCP process for stabilization must be conducted for each waste stream requiring stabilization (e.g., Class B waste, Class C waste) prior to burial. Qualification shall be conducted in accordance with the stabilization testing recommendations of the 'Waste Form Technical Position" Revision 1, January 1991, U. S. NRC. The SPCP stabilization qualification testing is typically performed by a CNS approved vendor.
- e.
Scaling Factors:
As permitted by 10CFR61.55, waste stream radionuclide scaling factors based on certain gamma emitting nuclides (e.g., Co-60, Cs-137) may be used to calculate the concentrations of difficult-to-measure radionuclides.
Scaling factor determinations are made using a detailed radionuclide analysis of waste stream samples conducted by a CNS approved vendor.
- f.
Reports:
The volume, curie content, principle nuclides, type of waste and solidification agent for the wet wastes SOLIDIFIED at Cooper Nuclear Station are documented in the Cooper Nuclear Station Radioactive Effluent Release Report. This information is listed in the format outlined in Revision I of Regulatory Guide 1.21.
F.
Dewatering:
- a.
General The CNS Dewatering PCP (DPCP) utilizes dewatering equipment and disposable waste containers to dewater solid wet waste streams produced from the liquid radioactive waste treatment system or from chemical decontamination resins.
Compliance with the DPCP ensures that the resultant waste meets the free standing liquid criteria set forth in I CFR61. The DPCP is not intended to be a substitute for the CNS Operations Manual nor does the process meet the waste stability form requirements of I OCFR61. Stabilization/solidification of the dewatered waste is provided by High Integrity Containers (HICs) and/or Engineered Concrete Barriers.
Solidification of the dewatered waste can also be accomplished by dewatering the waste in metal liners.
Solid wet waste streams produced from the liquid radioactive waste treatment system and chemical decontamination resins can be dewatered utilizing the CNS 8/7/00 I
CNS PCP DPCP. The liquid radioactive waste treatment streams include condensate resins, spent resins, reactor water cleanup resins and waste sludges. When regulations require solidification or stabilization, these waste streams can be dewatered in HICs and burial liners per DPCP requirements. The waste processing is performed with the Nuclear Packaging, Inc. dewatering system equipment installed at CNS, under the direct supervision of CNS personnel utilizing CNS approved procedures. The design and arrangement of the CNS dewatering system components are based on maintaining the operator radiation exposure ALARA. ALARA considerations and radiological controls are also incorporated into the system operating procedures.
After the waste has been processed in its burial liner the waste classification and type are determined so that the waste may be shipped to a licensed burial site in accordance with appropriate NRC, DOT, state and burial site regulations.
- b.
Operation - General:
The solid wet radioactive waste dewatering process is operated on a batch basis.
A batch consists of all the resulting dewatered liners processed from the contents of a waste source such as a phase separator. The batch is complete when the waste source is empty or the batch process is terminated due to radioactive waste system operational constraints.
Solid wet waste sources which are typically dewatered on a batch basis include the spent resin tank, waste sludge tank, condensate phase separators, reactor water clean-up phase separators, decontamination resins, or a combination of these waste sources.
Prior to dewatering the solid wet waste, verification is made that the waste being dewatered is characterized in accordance with the NUPAC Services Division, Inc. Waste Characterization Form. Selection of the burial liner (steel liner or HIC) utilized in the dewatering process is based upon the classification and type of waste being processed.
The CNS dewatering system consists of a dewatering waste container, a dewatering pump, an off-gas vent unit, a container level indicator, a waste fill head, a water separator with water chiller unit, an air blower, a relative humidity instrument, a control panel, and interconnecting piping and valves.
After wet radioactive waste is charged into a waste container, dewatering is achieved with continuous suction on the waste container by the dewatering pump.
Various types, numbers, and configurations of filters are used within the waste container to retain spent resin and filter precoat materials. Water removed from the waste container is returned to the liquid radwaste system.
The dewatering pump is operated for given time internals in accordance with the NuPac Services Division, Inc. Waste Characterization Form. The pumping time may range from eight to sixteen hours depending upon the waste stream and waste container. After most of the free water in the waste container has been removed, drying air is continuously recirculated in a loop from the air blower to the waste container through the water separator to remove any residual free water in the waste container.
The dewatering system is provided with temperature instrumentation which is interlocked to automatically shut down the dewatering process on high air temperature due to the potential for exothermic reactions during ion exchange resin dewatering. The waste container is considered dewatered when at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of pumping time has been completed and, the relative humidity in the recirculating drying air meets the acceptance criteria specified in the NUPAC Services Division, Inc. Waste Characterization Form. A relative humidity instrument and monitor are provided to remotely and continuously monitor the waste container outlet air. This instrument is used to establish the positive end point to the dewatering process and the point at which the free standing liquid criteria of IOCFR61 is met. Filled waste containers which do not meet the dewatering end point criteria are retained until a satisfactory corrective disposition is determined.
Stability or solidification of the waste is provided by the waste container and/or engineered barriers such as concrete vaults.
No airborne or liquid radwastes are released to the environment from the dewatering operation. The dewatered liquid radwastes are routed to the liquid radioactive waste treatment system and resin drying air is vented through a HEPA system. The dewatering system is designed to prevent the uncontrolled release of radioactive materials by monitoring liquid levels in the waste container by a level indicator. During the waste filling operation, the operator visually monitors the CNS PCP 8/7100 I
waste transfer process by observing the dewatering control panel video monitor and the local area radiation monitors. Potential inadvertent spills and overflows are contained in the Augmented Radioactive Waste Building and can be routed back to the liquid radioactive waste treatment system via floor drains.
- c.
Operation - Specific:
The CNS Dewatering System fillhead is installed on a waste container compatible with the dewatering fillhead. The waste container typically has four filter levels. The filters provide the flow path from the water container, through the fillhead, to the dewatering pump. The fillhead is fitted with connections to transfer solid wet waste from the Waste Transfer System to the waste container. The fillhead also contains a remote video camera for internal monitoring of the waste container, dewatering pump suction connections, electrical service connections for the waste container level indication system, a service air supply to provide cooling, a spray ring header for flushing, a pressure switch to monitor container pressure, and connections for the waste container temperature indicating system.
Once the fillhead is installed and the bottom filter level dewatering pump suction valve is opened, solid wet waste from the Condensate Phase Separator, Spent Resin Tank, or Reactor Water Cleanup Phase Separator is transferred as a homogeneously mixed slurry to the waste container. During the waste transfer the fillhead video monitor is used to verify that the waste container is being filled.
When the waste slurry reaches the second filter level, the dewatering pump is started. The dewatering pump is cycled on and off to maintain the water level in the waste container slightly above the level of the waste. When the slurry level reaches the third level of filters the second level dewatering pump suction valve is opened and the first level dewatering suction valve is closed. At this point a sample of the waste slurry is drawn for gamma isotopic analysis. As the waste slurry level reaches the fourth filter level the third level dewatering pump suction valve is opened. When the slurry level reaches the fill line deflector plate the fourth level dewatering pump suction valve is opened. Waste slurry transfer is continued until 87/100 I
CNS PCP the waste container is full. Demineralized water is then used to flush the residue in the waste transfer lines and fill head connections to the waste container. The dewatering pump is stopped, the second and third level dewatering suction valves closed, and the Waste Transfer System lineup secured. The dewatering cycle is then commenced by starting the dewatering pump.
When the fourth level dewatering pump suction line vacuum fluctuates rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the fourth level dewatering suction valve is closed and the third level suction valve is opened. As the third level dewatering pump suction line vacuum begins to fluctuate rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the third level dewatering suction valve is closed and the second level dewatering suction valve is opened. Similarly, as the second level dewatering pump suction line vacuum begins to fluctuate rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the second level dewatering suction valve is closed and the first level dewatering suction valve is opened. When the first level dewatering pump suction line vacuum fluctuates rhythmically with the dewatering pump and/or drops to approximately 5 inches Hg the first level dewatering suction valve is closed and the dewatering pump secured.
After the temperature of the radioactive waste in the waste container is measured and recorded the drying cycle is commenced. The drying cycle utilizes a blower to blow warm dry air through the waste to evaporate the residual water remaining in the waste container. The warm moist air which exits from the waste container is pumped through an entrainment separator tank where refrigeration coils condense the water vapor and the entrained water is removed. The water collected in the entrainment separator tank is pumped to the radioactive waste floor drain system.
The dehumidified warm air is recirculated through the waste until the endpoint humidity and minimum recirculation dry time requirements of the NuPac Waste Characterization Form are met.
The drying cycle begins by closing the fillhead waste isolation valve, aligning the blower to the waste container and starting the dewatering system chiller. The dewatering pump and blower are started and the blower inlet pressure, blower C8/7/00 I
CNS PCP discharge pressure, dewatering pump suction pressure, blower inlet temperature, blower discharge temperature, waste container temperature parameters are then monitored and recorded each hour. After 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of blower operation the waste level in the waste container is checked. Due to volume reduction inherent with the drying cycle, additional waste may have to be added to the waste container. To add waste, the drying cycle is secured and waste is added per the filling and dewatering cycles. Once additional waste is added the drying cycle is resumed. Several iterations of blower operation and waste addition may be required.
When 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of blower operation have been completed, and waste addition is not required, the Relative Humidity Monitoring System is placed in operation. This system monitors the relative humidity of the air stream from the waste container before the air enters the entrainment separator tank. The sample tubing is wrapped with heat tape and insulated to prevent condensation in the sample lines. The air sample flows through a filter/separator and then into the sample chamber. Air flow is maintained and monitored with a flow meter. A General Eastern 100DP Dew Point Hygrometer measures the system air temperature and the system air dew point temperature.
From this data, the percent (%) relative humidity of the entrainment separator inlet air is determined.
The air temperature and dew point temperature of the air stream at the inlet of the entrainment separator tank is then monitored and recorded hourly in addition to the temperature and pressure parameters previously noted. The percent relative humidity of the air stream is calculated using the air temperature and dew point temperatures. The drying cycle is considered complete when the percent relative humidity of the air stream reaches the end point relative humidity determined by the NuPac Services Division, Inc. Waste Characterization Form and the NuPac Services Division, Inc. Humidity Endpoint Graph and, when at least eight hours of blower operation has been completed. Once the drying cycle is successfully completed the CNS Dewatering System is shutdown, the fillhead removed from the waste container and the container primary and secondary closure devices installed.
If the percent relative humidity endpoint can not be attained the waste container is retained for corrective disposition. Waste containers which have been successfully 181700 I
CNS PCP dewatered are classified and shipped to a licensed burial site in accordance with the applicable sections of 10CFR61, 10CFR71, and 49CFR Parts 171 through 178.
- d.
Qualification:
Qualification of the DPCP process for dewatering was conducted in accordance with reference 19. Qualification of High Integrity Containers used for stabilization was conducted in accordance with references 17 and 18.
- e.
Scaling Factors:
As permitted by 10 CFR61.55, waste stream radionuclide scaling factors based on certain gamma emitting nuclides (e.g., Co-60, Cs-1 37) may be used to calculate the concentrations of difficult-to-measure radionuclides. Scaling factor determinations are made using a detailed radionuclide analysis of waste stream samples, by a CNS approved vendor.
- f.
Reports:
The volume, curie content, principle nuclides, and type of waste for the wet wastes dewatered at Cooper Nuclear Station are documented in the Cooper Nuclear Station Radioactive Effluent Release Report. This information is listed in the format outlined in Revision I of Regulatory Guide 1.21.
G.
Vendor Process Control Pro-grams, Stabilization Processes, and High Integrity Contain Qualifications:
Vendor supplied Process Control Programs, Stabilization Processes and High Integrity Containers which have NRC approved Topical Reports or, which are approved in Low-Level Radioactive Waste Disposal Facility licenses, can be utilized to process radioactive waste at CNS.
H.
District Initiated Changes
- 1.
Shall be submitted to the Commission by inclusion in the Radioactive Effluent Release Report for the period in which the change(s) was made effective and shall contain:
- a.
Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information;
- b.
A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and
- c.
Documentation of the fact that the change has been reviewed and found acceptable by the SORC.
- 2.
Shall become effective upon review and acceptance by the SORC.