ML20206J266
| ML20206J266 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/22/1999 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20206J237 | List: |
| References | |
| PROC-990422, NUDOCS 9905120171 | |
| Download: ML20206J266 (170) | |
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COOPER NUCLEAR STATION I OFFSITE DOSE ASSESSMENT MANUAL
-ODAM- l LIST OF EFFECTIVE PAGES Egge Date Eage Date Page Date i 10/15/96 40 12/29/88 Appendix D (Cont.)
ii 1/27/89 41 12/29/88 1 10/15/96 42 Original DI.0-1 8/15/98 2 10/15/96 43 Original 3 Original 44 05/16/94 D2.0-1 8/15/98 4 10/15/96 45 Original 5 10/15/96 46 05/16/94 D3.0-1 8/15/98 6 10/15/96 47 Original l 7 10/15/96 48 Original D3.1-1 8/15/98 l 8 10/15/96 49 Original D3.1-2 8/15/98 9 10/15/96 50 Original D3.1-3 8/15/98 10 10/15/96 51 12/29/88 .D3.1-4 4/22/99 11 10/15/96 52 Original D3.1-5 8/15/98 l l
12 10/15/96 53 Original D3.1-6 8/15/98 13 12/29/88 54 Original D3.1-7 8/15/98 14 12/29/88 55 Original D3.1-8 8/15/98 I
15 05/16/94 56 Original D3.1-9 8/15/98 16 12/29/88 57 Original D3.1-10 8/15/98 17 Original D3.1-11 8/15/98 18 Original Appendix A:
19 Original (Deleted) 1/27/89 D3.2-1 8/15/98 )
20 10/15/96 D3.2-2 8/15/98 21 Original Appendix B: D3.2-3 8/15/98 22 Original Pages B -I through D3.2-4 8/15/98 23 10/15/96 B-7 Original D3.2-5 8/15/98 24 10/15/96 D3.2-6 8/15/98 25 10/15/96 Appendix C: D3.2-7 8/15/98 j 26 03/06/96 C-1 Original D3.2-8 8/15/98 1 27 12/29/88 C-2 5/16/94 D3.2-9 8/15/98 l 28 12/29/88 C-3 03/06/96 D3.2-10 8/15/98 j 29 1/27/89 C-4 05/20/97 D3.2-11 8/lS/98 '
30 12/29/88 C-5 03/06/96 D3.2-12 8/15/98 31 1/27/89 C-6 03/06/96 D3.2-13 8/15/98 32 12/29/88 C-7 05/20/97 D3.2-14 8/15/98 33 12/29/88 C-8 8/30/90 D3.2-15 8/15/98 34 05/16/97 C-9 5/16/94 D3.2-16 8/15/98 35 1/27/89 C-10 05/20/97 36 12/29/88 D3.3-1 8/15/98 37 1/27/89 Appendix D: D3.3-2 8/15/98 38 12/29/88 D3.3-3 8/15/98 39 12/29/88 lii 8/15/98 D3.3-4 8/15/98 NOTE: Original refers to page in effect on July 1,1986, when the ODAM was implemented.
9905120171 990503 ' age 1 of 2 April 22,1999 PDR ADOCK 05000298 R PDR
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J COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL
-ODAM-LlST OF EFFECTIVE PAGES (CONTINUED)
Eage Date Eage Date Pane Date Appendix D (Cont.) DS.4-2 8/15/98 B53-1 8/15/98 D33-5 8/15/98 D5.5-1 8/15/98 B5.4-1 8/15/98 D33-6 8/15/98 D33-7 8/15/98 iv 8/15/98 B5.5-1 8/15/98 D33-8 8/15/98 D33-9 4/22/99 B3.1-1 8/15/98 D33-10 8/15/98 B3.1-2 8/15/98 D33-11 2/09/99 B3.1-3 8/l5/98 D33-12 8/15/98 B3.1-4 8/15/98 D33-13 8/15/98 B3.1-5 8/15/98 D33-14 8/15/98 D33-15 8/15/98 D33-16 8/15/98 B3.1-6 8/15/98 D3.3.17 8/15/98 B3.1-7 8/15/98 B3.1-8 8/15/98 D3.4-1 8/15/98 B3.1-9 8/15/98 D3.4-2 8/15/98 B3.2-1 8/15/98 D3.5-1 8/15/98 B3.2-2 8/15/98 D3.5 2 8/15/98 B3.2-3 8/15/98 B3.2-4 8/15/98 D4.1-1 8/15/98 B3.2-5 8/15/98 D4.1-2 8/15/98 B3.2-6 8/15/98 D4.1-3 8/15/98 B3.2-7 9/29/98 D4.1-4 8/15/98 D4.1-5 8/15/98 B33-1 8/15/98 D4.1-6 8/15/98 B33-2 8/15/98 D4.2-1 8/15/98 B3.4-1 8/15/98 D4.2-2 8/15/98 B3.5-1 8/15/98 D43-1 8/15/98 B4.1-1 8/15/98 DS.1-1 8/15/98 B4.2-1 8/15/98 D5.2-1 8/15/98 D5.2-2 8/15/98 B43-1 8/15/98 D53-1 8/15/98 B5.1-1 8/15/98 D5.4-1 8/15/98 B 5.2-1 8/15/98 NOTE: Original refers to page in effect on July 1,1986, when the ODAM was irnplemented.
Page 2 of 2 April 22,1999
COOPER NUCLEAR STATION OFFSITE DOSE ASSESSMENT MANUAL FOR GASEOUS AND LIQUID EFFLUENTS Eau 1.0 I ntrod uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l .
2.0 Liq u id E ffl ue n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 2.1 Radioactivity In Lio uld Wa ste . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -2 2.2 Aq ueou s Conce n tra tio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -2 2.3 Method of Establishine Alarm Setooints . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.3.1 Setooint for a Batch Release ............................................ -5 2.3.2 Setooint for a Continuous Release . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8-2.4 Radioactivity Concentration in Water Beyond the Site and Exclusion Area Boundarv . . . -10 2.5 Accumulated Personal Maxim u m Dose . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 12 2.6 Proiected Personal Maximum Dose ........................................... 1 l
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3.0 G aseous E ffl ue n t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . l 3.1 I n trod u ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 17 3.2 Radioactivity in Gaseous Effluent . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17-I 3.3 Main Condenser Air Eiector Noble Gas Monitor Alarm Setooint . . . . . . . . . . . . . . . . . . . . -19 l 3.4 Effluent Noble Gas Monitor Alarm Setooint . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -20 3.4.1 Setooin t Ba sed on Dose Rate . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -20 3.5 Noble Gas Gamma Radiation Dose Accumulated in Air ...........................25-3.6 Noble Gas Beta Radiation Dose Accumulated in Air . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 3.7 Dose Due to Iodine and Particulates in Gaseous Effluents * . . . . . . . . . . . . . . . . . . . . . . . . 30 3.7.1 G ASPAR Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31-3.7.2 Alte rnate Me thod . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 2 3.8 Dose to a Person from Noble Gases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -35 3.8.1 Ga mma Dose to Total Body - GASPAR Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . i ,
3.8.1.1 Al te rn a te Me t h od . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 G-3.8.2 Dose to S kin - GAS PAR Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -3 7- !
3.8.2.1 Alte rn a te Method . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 37-3.9 Projected Organ Dose Due to Gaseous Effluent . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3.10 Dose Rate Due to Tritium. Iodines. and Particulates in Gaseous Effluents . . . . . . . . . . . . 39-4.0 Dose Commitment From Releases Over Extended Time . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 42-4.1 Rele a se s D u rin e A O u a rter . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -42 4.2 Rele a se s D u ri n e 12 Mo n t h s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -44 4.2.1 Calcula ted Do ses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -45 4.2.2 Environ mental Measure m ents . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 7-5.0 Radiological Environmental Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -49 5.1 Environ m ental Sampline Procra m . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49-
.aPENDIX A -(Deleted)
APPENDIX B - Reference Meteorological Data APPENDIX C - Environmental Radiation Monitoring Program Sample Types and Sample Station Locations APPENDIX D ODAM Specifications e
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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 maxunum 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.
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l 2.0 Liquid Efnuent 2.1 Radioactivity In Liquid Waste The concentration of radionuclides in liquid waste is determined by sampling and analysis in accord with Table D3.1.11, Radioactive Liquid Waste Sampling and Analysis. Alternatively, pre release analysis of the radioactivity concentration in liquid waste required by DSR 3.1.1.1 may be done by gross p-y counting provided an effluent concentration beyond the site and exclusion area boundary for unidentified emitters,1 x 10 4pCi/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.
2.2 Aoucous 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 Cg =C,}F 2
where C, = concentration of radionuclide i in liquid radwaste released ( Ci/ml)
C, = concentration of radionuclide iin the receiving stream (pCi/ml)
F = release rate ofliquid radwaste (ml/sec)*
F, = dilution Gow of receiving stream of water (ml/sec)*
- F , F , and F, may have any convenient units of flow (i.e., volume / time) provided the units of i
all are identical.
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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 F .
This method of estimating concentration of radionuclide iin 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:
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l For Sr 89, Sr 90 and Fe-55 l
Czi= N Vdq where Czi = concentration of radionuclide iin the receiving stream ( cilml)
Cqci = concentration of radionuclide i in the quarterly composite sample (pci/ml)
Vrq = volume ofliquid radwaste discharged during the quarter (ml)
Vdq = total volume of dilution flow corresponding to the time when Vrq was discharged for the quarter (ml) 1 l
For all other nuclides Czi = #A Vdm where Czi = concentration of radionuclide i in the receiving stream (pci/ml)
Cmei = concentration of radionuclide i in the monthly composite sample (pci/ml)
Vrm = volume ofliquid radwaste discharged during the month (ml)
Vdm = total volume of dilution flow corresponding to the time when Vrm was
' discharged for the month (ml) 2.3 Method of Establishine Alarm Setooints The liquid waste emuent 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.
l The alarm setpoint for the liquid emuent 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 emuent line and the service water emuent line through which radioactive emuent is, or may l be, eventually discharged into the discharge canal.
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The alarm setpoint calculation for each liquid effluent monitor is based upon measurement according to Table D3.1.11 of radioactivity in a batch of liquid to be released or in the continuous aqueous discharge. Alternatively, the alarm setpoint l
l may be based upon gross p-y activity analysis of the liquid waste provided the effluent concentration beyond the site and exclusion area boundary for unidentified emitters, 1 x 104 pCi/ml, is observed.
In any case, a monitor may be set to alarm or trip at a lower activity concentration than the calculated setpoint. j 2.3.1 Setnoint for a Batch Release l
A sample of each batch of liquid radwaste is analyzed for I-131 and principal gamma emitters, or for total activity concentration prior to release. The ratio, FMPC,, of the i 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
"' identified FMPC* = 1 MPC, l
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where FMPC,, = fraction of effluent concentration beyond the site and exclusion area boundary in batch derived from activity measured prior to release.
C ,, = concentration of radionuclide a (including I-131 and principal garuma emitters) in batch sample taken prior to release (pCi/ml).
MPC,= effluent concentration beyond the site and exclusion area boundary of radionuclide i per 10 CFR Part 20, Appendix B, Table 2, Column 2 (pCi/ml)
When FMPC3 , is derived from analyses identifying iodine and principal gamma 1
emitters only, the value FMPC 3, 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:
FMPC, = FMPC,, + E, Previous quarterly average of the fraction of the effluent concentration in the where E, = discharee canal due to I-131 and crimarv camma emitters Previous quarterly average of the fraction of the effluenc concentration in the discharge canal due to all radionuclides in batch releases.
A reference value of E,, 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 l
valve closure setpoint is determined with the equation:
1 S=
^ '
g+B EMPC, Ey 10/15/9G j L_-- __ - _ ___ _ ____ ________ _ ____ _ ______ _ _ _ _ ___ _ _ ___
where S = radiation monitor alarm setpoint (cpm or pCi/ml)
A = counting rate (cpm /ml) or activity concentration (pCi/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 pCi/mlor pCi/ml per pCi/ml) l l Fsi = maximum flow in the batch release line (gal / min)**
F,3
= minimum flow in the discharge canal (gal / min)**
Bkg = monitoring instrument background (epm or pCi/ml)
Note that A+FMPC, represents the counting rate of a solution having the same radionudido distribution as the sample and having the maximum permissible concentration of that mixture.
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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 l exclusion area boundary in the batch is:
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l C*
FMPC* = 1 x 10-8 where Cg = gross or total radioactivity concentration in batch sample taken prior to release ( Ci/ml) 1 x 10 4 = effluent concentration beyond the site and exclusion area boundary of unidentified radionuclides (pCi/ml) l
- A equals E Cy ifisotopic analysis was performed or C g if gross activity analysis was I performed.
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- Any suitable but identical units of flow (volume / time).
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l The value of FMPC9 computed with this expression is substituted in the preceding equation to calculate the setpoint. J I
l 2.3.2 Setnoint for a Continuous Release Continuous aqueous radioactive discharges are sampled and analyzed according to the schedule in Table D3.1.1 1. The ratio FMPC,,,, 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'" = iden tifled 1 HPC, where FMPC,,, = fraction of efIluent concentration beyond the site and exclusion area boundary in continuous release based upon activity measured in weekly composite sample (s).
1 Ca = concentration of radionuclide i (including I 131 and principal gamma emitters) in weekly composite sample (s)
(pCi/ml) ,
When FMPC,is derived from anilyses ofI 131 and principal gamma emitters, it may j l be adjusted to account for radionuclides measured in the monthly and quarterly ]
! composite sample but not measured prior to release. Adjustment for radionuclides I
measured in monthly and quarterly composite samples but not in weekly composite I samples is given by the equation FMPC, = FMPC,,, + E, i
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Quarterly average fraction of the effluent concentration in the discharge canal 1
due to I-131 and primary gamma emitters measured in weekly where E, = composite anmole of continuous releases durine orevious auarter 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 E,, 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 l
equation A
- S= g+B FMPC, F,,
where A = counting rate (cpm /ml) or activity concentration (pCi/ml) of weekly composite sample in the laboratory.
l Terms g, Fsi, F3,, 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 efDuent concentration beyond the site and exclusion area boundary in a sample of the release is:
C' EMPC* =
l 1 x 10 8
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where C, = gross or total radioactivity concentration in continuous aqueous release (pCi/ml) l 1 x 10-8 = effluent concentration beyond the site and exclusion area l
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 t
i 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 Bevond 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 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.
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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 i 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:
EMPC =
1 E 1 E Nrf 378 5 ( TE-TB) k E2, i 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 s 1) 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)
F2, = 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)
Qu = quantity of radionuclide i represented by sample k which is released as an effluent within the time boundaries TB and TE (uCi)
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MPC,= maximum permissible concentration beyond the site and exclusion area i boundary of radionuclide i per 10 CFR Part 20, Appendix B, Table 2, Column 2 ( Ci/ml)
The data used to compute FMPC are measured by the radioactive hquid sampling and analysis program described in Table D3.1.1 1.
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2.5 Accumulated Personal Maximum Dose l
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 Annual Operating Report Radioactive EfDuents Docket Number 50 298.
- With quality factor for Tritium reduced from 1.7 to 1.0 per ICRP.
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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 efs, the average flow in the discharge canal during the time ofinterest.
o Dilution factor for the effluent.
Drinking water: s5 (for LADTAP variable Dilution factor)
Fish *: s5 (for LADTAP variable Dilution factor)
- Fishing - Seasonal variation: Consumption of fish is evaluated from April 1
l through November.
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l Alternatively, the accumulated dose commitment may be calculated in the following 1
way:
t i EE F1 AD,,, = 3. M x W g 3 A,,,, C,, M, y i
s 2; k D,, = E g bD,,,
l where AD,a = the' dose commitment (mrem) to organ n of age group a due to !
the isotopes in a release represented by analysis k, where the l 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 1
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.
D, = 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).
A,,a = 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 min / gal 3.785 x 10'8 = 3785 ml/ gal x 104 Ci/pCi Ca = the concentration of radionuclide i in the undiluted liquid waste to be discharged ( Ci/ml), i.e., in the sample k Atg = elapsed time of release represented by sample k during which radionuclide i is discharged at concentration C , i.e., the duration of the release represented by sample k (minutes)
(F/i2
/F ) = the quotient of the release flow, F , and the dilution flow, F ,
i during the release represented by sample k l
l 12/29/88 1
1 Pathway-to-dose transfer factors, A, 3, 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, l
Revision O. 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 l and exclusion area boundary near the discharge canal, F = 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 1
of the Public via drinking water will be assessed on the basis of water assumed to be 1
l taken from the river three miles downstream. At that location, F is conservatively assumed to be F, = 5F,. Variables F i, F , and F, are defined in Section 2.2.
2.6 Proiected 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 proiected over a quarter at least one time during every 31 days if 15 05/16/94
i 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 l 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 l
l projected doses during the current quarter. Doses will be calculated in accordance with Section 2.5.
As an alternative, the quarterly dose commitment to the total body and most exposed organ may be projected by using the equation P,, = 91 D,,
where P., = projected dose commitment (mrem) to organ n (including total j i
body of age group a during the current quarter l 1
91 = number of days in a quarter !
i i X = number of days to date in current quarter l D, = dose commitment during the quarter-to-date (mrem) based upon results of aqueous effluent sampling and analyses available to 1
I date during the quarter 12/29/88
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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 31. 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 D3.2.31 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.
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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 efDuent gas samples in accordance with DLCO 3.2.3, Table D3.2.31. 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 O = 2. 8 x 10' 5 F 9
where Q = total radioactive noble gas release via a gaseous effluent stream during a given time interval (pCi)
-18
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t l N = not counts accumulated during the time interval g = effluent noble gas monitor counting rate responso ,
1 l t 1 cpm l ,
r pCi/ca',
l Fp = gaseous effluent stream discharge rate (cfm) 8 l 2.8 x 10' = conversion constant (cm /ft')
l 3.3 Main Condenser Air Eiector Noble Gas Monitor Alarm Setooint 1
1 A noble gas activity monitor is provided to measure gross gamma activity in gases at 1
l the main condenser air ejector. The monitor includes an alarm that is set to report l when the gamma radiation levelin gas discharged by the main condenser air ejector indicates the gross radioactivity discharge rate exceeds 1 Ci/sec.
I The alarm setpoint is determined with the relation 1
h ^D S = 2120 + Bkg F
l l l
where S = the alarm setpoint (mr/hr) h = monitor response to activity concentration of SJAE offgas being monitored (mr/hr per (pCi/cm*))
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 (unitiess)
Ci 1.0E6pCl 1 ft 3 60 sec 2120 = 1 x x x Sec. Ci 28317 cm 3 min.
l j -19
3.4 Effluent Noble Gas Moaitor Alarm Setooint DSR 3.3.2.9 requires an alarm setpoint to be determined for each radioactive noble gas effluent monitor. Each setpoint is derived to cause the alarm to report when the dose equivalent rate beyond the site and exclusion area boundary due to radioactive noble gas in gaseous effluent exceeds a limit in DLCO 3.2.1.a. Each noble gas activity monitor included in Table D3.3.21 except the main condenser air ejector off gas monitor is set to initiate alarm at or below the derived setpoint.
For the purpose of deriving a setpoint, the distribution of noble gas radionuclides in an effluent stream is determined as described in Section 3.2.
3.4.1 Setnoint Based on Dose Rate The alarm setpoint of a radioactive noble gas effluent monitor may be calculated on the basis of whole body dose equivalent rate beyond the site and exclusion area boundary. A setpoint of a monitor of an elevated release, e.g., from the stack, may be calculated with the equation.
10/15/96
E C, hP A S = 1. 0 6 , + Bk f D,C, DF*, 1
- i The setpoint of a monitor of a ground-level or split-wake release, e.g., from the turbine building vent or the AOG building, may be calculated with the equation E C, hP i S = 1. 0 6 , + Bkg 5E D,C, DF,i 1 g i where S=
the alarm setpoint (cpm, mr/hr, or pCi/cm 8) h = monitor response to activity concentration of effluent being monitored, (cpm per pCi/cm8 , mr/hr per pCi/cm8 , or pCi/cm8 per pCi/cm8)
C, = relative concentration of noble gas radionuclide i in efHuent at the point of monitoring (pCi/cm8)
X/Q = atmospheric dispersion from point of ground level or split-wake release to the location of potential exposure (sec/m8) s DF, = factor converting elevated release rate of radionuclide i to total body dose equivalent rate at the location of potential exposure
mrem yr.gCJ see DF[ = factor converting ground. level of split-wake release of radionuclide i to the total body dose equivalent rate at the location of potential exposure mrem yr.gCi, m
f = flow of gaseous efDuent stream, i.e., flow past the monitor (ft*/ min)
Bkg = monitoring instrument background (epm, mr/hr, or pCi/cm8) 1.06 = 500 mrem /yr x 60 sec/ min x 35.3 8ft /m 8
x 1.0 ms /1.0 x 108 cms P = fraction of allowable limit representing a chosen margin of conservatism in the setpoint (unitiess)
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.
The atmospheric dispersion and the dose conversion factor, DF,8 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 j 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 4 or Split-Wake ERP Elevcted W 1,800 8.2 x 10-8 The applicable dose conversion factors, DF', and DFi for deriving setpoints are in Table 3 3.
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3.5 Noble Gas Gamma Radiation Dose Accumulated in Air ;
i 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.
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l The distribution of radioactive noble gases in gaseous releases and the quantity discharged during an interval ofinterest are determined as described in Section 3.2.
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 Annual Operating Report Radioactive Effluents Docket Number 50-298.
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 Fraction of year milch animals are on pasture: 0.5 (April - Sept.)
o Parameters for specified locations:
Distance and compass direction (in degrees) from station to specific location.
- Noble gas (gamma and beta) doses to air, 1106 meters NNW, 335 degrees, site boundary.
- Most exposed Residence, 1770 meters West, 270 degrees, or 1448 meters NW,305 degrees.
- Milch cow,5634 meters NW,326 degrees.
- Quality factor for Tritium reduced from 1.7 to 1.0 per ICRP.
03/06/96
i I
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 I
with the equation:
f \ f \
f i D=E3
- E
- X Ocs,
- Ayes, i "i g' "
- AYv, s i r >
where D = noble gas gamma dose to air (mrad)
Q = E AQ i= cumulative release of noble gas nuclide i from stack (pCi).
Ay = factor converting unit noble gas stack release to ground level air dose 9 from overhead plume gamma radiation (mrad /pCi).
Ay,8
= factor converting time integrated, ground level concentration of noble gas to air dose from gamma radiation mrad '
pCi sec
( m;3 27 12/29/88
Qu = .E AQcvi = cumulative release of noble gas nuclide i from building time vents ( Ci).
[X) long term average atmospheric dispersion factor for a ground level or
=
l- l
\Q/,, split wake release (sec/m'). 1 DSR 3.2.2-1 is satisfied by calculating the noble gas gainma radiation dose to air !
beyond the site and exclusion area boundary identified in Figure 3-2 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 ,
4 is X/Q = 3.4 x 10 sec/m8 at the NNW eite boundary. Appropriate values ofYA i 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 3.2.21 requires the air dose to be calculated on a cumulative basis.
The radioactive noble gas distribution and activity discharged are determined as described in f 3.4 herein.
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- Onsite meteorological data for the period July 1,1976, to June 30,1977, which was used in the Cooper Station Demonstrated of Comoliance with 10 CFR 50. Appendix I.
revision 1, January,1978.
12/29/88
I The beta rad.~ation 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 i l
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 l l
boundary as a consequence of noble gas released from the station may be calculated I with the equation X X' D=E'O 1 c
4Q
+ Q%
- A>,) l s
a 0a's Where D
= noble gas beta dose to air (mrad)
X = long-term average atmospheric dispersion factor for stack Qa 8 releases (sec/m )
Ap. = 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 the location identified in Figure 3-2 and on the basis of reference atmospheric dispersion assuming continuous gaseous discharge.
At that location, the reference atmospheric dispersion factors are:
X = 1.2 x 104 sec/m8 at the NNW site boundary Q.
X = 3.4 x 104sec/m 8 Q,
Beta radiation-to-air dose conversion factors, Ap, for noble gas radionuclides are listed in Table 3-4.
I
- Quality factor for Tritium reduced from 1.7 to 1.0 per ICRP.
1/27/89
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 halflives greater than eight days in gaseous effluents released to the area beyond the site and exclusion area boundary cause no more than 7.5 mrem to any organ of a member of the public during any calendar quarter or 15 mrem to an organ of a member of the public during any calendar year. DSR 3.2.3.2 requires the dose to be calculated at least once every 31 days.
Radionuclides other than noble gases or tritium in gaseous effluents that are measured by the sampling and analysis program described in Table D3.2.3-1 are used as the release term in dose calculations. Airborne releases are discharged either via the stack (ERP) as an elevated release or via building vents and treated as a ground level or split wake release. For each of these release combinations, samples are analyzed weekly, monthly, quarterly, or for a specific release according to Table D3.2.3-1.
Each sample provides a measure of the concentration of specific radionuclides, C,, in gaseous effluent discharged at flow, F,, during a time increment At. Thus, each release is quantified according to the relation:
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- The dose to any organ of a person arising from radioactive iodine-131, iodine 133, and radioactive materialin particulate form having half-lives greater than eight days. Noble gases not considered.
12/29/88
L l
0 0 ), = C,, F,, h t, 1
l E
01k *j C,, F,, bt, where Qa = the quantity of radionuclide i released in a given effluent stream based on analysis k (Ci)
Ca = concentration of radionuclide i in gaseous effluenk identified by analysis k (Ci/m 8) or (pCi/cm 8) l F,; = effluent stream discharge rate during time increment At(m 3
/sec) 8 At3 = elapsed time in increment j during which radionuclide i at concentration Ca is being discharged (sec)
I 3.7.1 GASPAR Method A person may be exposed directly to an airborne concentration of radioactive material discharged in efnuent 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, Qa, of radioactive material other than noble gas will be calculated in accordance with Regulatory Guide 1.109, Revision 1, utilizing USNRC Computer Code GASPAR.* l l
l l
The GASPAR program is routinely used for calculating radiological dose assessments. !
l Site specific parameters input to GASPAR are listed in Section 3.5.
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- Quality factor for Tritium reduced from 1.7 to 1.0 per ICRP.
1/27/89 l
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1 3.7.2 Alternate Method Alternatively, the dose commitment to a person beyond the site and exclusion area boundary associated with a gaseous release, Qa, of radioactive material other than noble gas may be calculated with one of the appropriate following equations l release via the stack:
/ i f i E Xd +EE D D ,,,, = Q ,,, 3 %,,, g 3 M ,,,, -
L A cs \ 0 1 bse release via a vent:
f i f i E Xd EE D anvk "O ikv i ani ei eani
~
\ }) bv \ 0)cve wher e D,,,, = 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 D3.2.31.
=
D.. the dose commitment from a vent release (mrem)
TA,a = factor converting airborne concentration of radionuclide i to dose commitment to organ n of a person in age group a f i mrem
, (Ci sec) /m3 ,
TG,,a = factor converting ground deposition of radionuclide i to dose commitment to organ n of a person in age group a exposed via environmental pathway e (mrem /Ci/m')
(D/Q) = relative deposition factor (m ')
12/29/88
(Xd/Q) = depleted atmospheric dispersion factor (mci /m8per 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 D3.2.31, become available.
The dose is accumulated in the following way.
The dose accumulated as a result of stack discharge is l
D,,, = E l
y D,,,, + E, D,,,,
+ E D,,,,
q and the dose accumulated as a result of a vent discharge is D,,, = E y D,,,, + E,D,,,,
+ E D,,,,
q Doses committed during the same time period due to discharges from the stack and vents are additive, thus:
D,, = D,,, + E y D,,,
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) 12/29/88
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, A,,,3, are selected for use in calculating the dose.
The dose to a receptor at the location identified in Figure 3-2,1.1 milec 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:
f f i T
- D
- 8.1 x 10-8 sec/m3 - 4. 6 x 10-2 m -2 s O>s r Q) s f i f i
- 4. 4 x 10-7 sec/m 3 - 9. 5 x 10-1 m -2 s O>v rQ iy 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 where a real milch animalis located,3.7 miles northwest of the Station, are:
r 3 l
D
- 1. 2 x 10-1 m -2 l rQ i s l
f 1
- 3. 7 x 10-10 m -2 s Os v 40 CFR Part 190. When the dose due to gaseous efauent 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.
05/16/94
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),
l air grass cow-milk, or air grass-cow meat pathways is calculated with the appropriate )
one(s) of the equations:
for a stack release l
l XE '
D, ,,, = 3 Q,,, Ep % ,,,,
for a vent release D,,,, = XE3 Q,,,Ep % ,,,,
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 Bodv 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.
-35 1/27/89
l 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:
f f ) \
E X '
N DY = 3 g
Os*N,*O,\
cs es cv
~
Al cv vi j
where D = noble gas gamma dose to total body (mrem)
Py g = factor converting unit noble gas nuclide i in stack release to total body dose at ground level received from the overhead plume (mrem / Ci)
Py,, = factor converting time integrated, ground level concentration of noble gas nuclide i to air dose from gamma radiation r 3 l
. mrem pCi sec
, m, 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 l
resident exposed to maximal ground level noble gas concentrations (maximum X/Q) may be selected as the receptor. The location of the latter residence is identified in Figure 3-2. Values by Pyes, and Pyv, applicable at the location of the residence 1.1 miles west of the station appear in Table 3-5.
12/29/88 l
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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.*
l 3.8.2.1 Alternate Method Alternatively, the beta radiation dose to the skin of a member of the public due to j beta radiation from noble gas released from the Station may be calculated with the equation l
X X' DQ =,;E'Q,,, g + Q ,,, SQ, g CS CVj where Dp = noble gas beta dose to skin (mrem)
Sp, = factor converting time integrated ground level concentration of noble gas radionuclide i to skin dose from beta radiation mrem (pci sec) /m 3 Values of Sp, for noble gases are included in Table 3-5.
When the skin dose due to noble gas beta radiation is evaluated as required by DSR 3.4.1.1, the receptor selected is the nearby resident exposed most via all applicable exposure pathways together. Alternatively, the nearby resident exposed
.to maximal ground-level concentrations (maximum X/Q) may be selected as the receptor. The location of the latter resident is identified in Figure 3 2.
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- Quality factor Tritium reduced from 1.7 to 1.0 per ICRP.
1/27/89
The total dose to the skin from noble gases is approximately equal to the beta radiation dose to the skin plus the gamma radiation dose to the total body.
3.9 Proiected Organ Dose Due to Gaseous Effluent I
l DSR 3.2.5.1 and DSR 3.2.4.2 requires organ dose to a member of the public due to l
l radioactive material in air effluent be oroiected during each month in which radioactive materialis released in gaseous effluent without treatment. The purpose is to guide plant personnel in operating the EVTS aad 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 l projecting it for an entire 31 day time by employing the equation:
I PD - D X
where:
l PD = projected organ dose to a member of the public (mrem) l 1
31 = number of days over which dose is projected I 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).
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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 l
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:
1 1 10-6 EE X D,,, = g_w y 3 Q,,,
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For effluent from a ground level release point, i.e., a building vent, the equation is.
I l
f )
10*' EE X l
D,,, =
_w y 3 0,g, Dose rates from separate release points may be combined to give D,, = D,,, + E y D,,,
where D., = dose equivalent commitment rate to organ n of a person in age group a due to radioactive particulates, iodine, and tritium in airborne effluent that are inhaled (mrem /hr) l l
l l
12/29/88
D,,,,D ,
= dose equivalent commitment rate due to radioactive particulates, iodine, and tritium from an elevated release and a ground-level release respectively (mrem /hr)
Q.u,Qvu = 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)
TE = ending time of efBuent discharge TB = beginning time of effluent discharge TE-TB = efBuent averaging time (hr)
(X/Q),,(X/Q), = atmospheric dispersion from an elevated or a ground level release respectively to ground-level at the receptor (uCi/m' per uCi/sec)
TA,,,i = factor converting airborne concentration of radionuclide i to dose commitment to organ n of a person in a'ge group a and where e represents exposure via inhalation r ,
mrem s
(Ci sec) /m3 ;
10 4 = conversion,104 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 efDuent during the time increment represented by the sample.
12/29/88
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4.0 Dose Commitment From Releases Over Extended Time 4.1 Releases Durina 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 i 2.6.
- 2. total body and maximally exposed organ doses due to gaseous effluents
- other than noble gases and tritium as in 6 3.7.
- 3. doses to air offsite due to noble gas y as in f 3.5 and due to noble gas p as in f3.6.
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 1
D3.2.3 1.
- radioactive iodine 131, iodine-133, and radioactive material in particulate form, having halflives greater than eight days.
l l 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 1 i
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 l l
identified in the annual land use census and maximum ground level X/Q at the l 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. I Seasonal nppropriateness 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.
-4 3-
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\
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.
I 4.2 Releases Durina 12 Months The regulation governing the maximum allowable dose or dose commitment to a I 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.
l 05/16/94
I The dose to a member of the public which is due to exposure to radioactive material in liquid and gaseous effluents from the station are ordinarily calculated while the dose attributable to irradiation is evaluated with environmental radiation dosimetry.
The receptor of the dose is selected on the basis of the combination of applicable l
pathways of exposure to gaseous effluent identified in the annual land use census I 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 l 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 efHuent are calculated as described by the equation for:
- 1. Total body and maximally exposed organ doses due to liquid efHuent via drinking water and eating fish from the river as in f 2.0.
- 2. Total body dose due to noble gas y as in f 3.8.1.
- 3. Skin dose due to noble gas p as in f 3.8.2.
-4 5-
- 4. total body and maximally exposed organ doses due to gaseous effluents
- other than noble gases as in 6 3.7.
Aqueous radioactive material concentrations are estimated according to 6 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 l greater than eight days.
l 05/16/94
l f 4.2.2 Environmental Measurementa i
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. l l
The dose to a member of the public due to irradiation (external exposure to gamma l 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, l 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 l
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 efDuent l
l 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 l about 85 miles, radioactive liquid effluent contamination of potable water is not foreseen to be significant. The other liquid effluent pathway of potential signi6cance, via fish taken from the river, would be evaluated
-47
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 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.
i
-4 8-
5.0 Radiological Environmental Monitoring Program 5.1 Environmental Sampline Procram 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 approximato location.
A radiological environmental monitoring program, approved by the Nuclear Regulatory Commission (NRC) was initiated ut 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.
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Table 3 2 Computed Release of Radioactive Noble Gases In Gaseous Effluent From Cooper Nuclear Station Stack Release Plant Vents Release Nuclide (Ci/yr) Fraction (Ci/yr) Fraction Kr-83m 3.60E+01 8.38E 03 0 0 Kr-85m 6.50E+01 1.51E-02 7.10E+01 1.14E-02 Kr-85 2.00E+02 4.66E-02 0 0 Kr-87 2.13E+02 4.96E-02 1.33E+02 2.13E-02 Kr-88 2.13E+02 4.96E-02 2.33E+02 3.74E-02 Kr-89 1.00E+03 2.33E-01 0 0 Xe-133m 3.00E 00 6.99E-04 0 0 Xe-133 1.51E+02 3.52E-02 2.63E+03 4.22E-01 Xo 135m 7.20E+01 1.68E-02 6.96E+02 1.12E-01 Xe-135 2.64E+02 6.15E-02 1.06E+03 1.70E-01 Xe 137 1.20E+03 2.79E-01 0 0 Xe 138 8.77E+02 2.04E-01 1.41E+03 2.26E-01 Total 4294. 1.0 6233. 1.0 Releases computed by BWR-GALE for Cooper Station Base Case gaseous radwaste treatment.
The release rate (Ci/yr) is included only to show the basis of the radionuclide distribution. To estimate the concentrations of radionuclides in a sample in which only the total radioactivity has. been measured, multiply the total activity concentration by the fraction of respective radionuclides listed above.
Table 3-3 Dose Conversion Factors for Deriving Radioactive Noble Gas Effluent Monitor Setpoints Radionuclide Factor Df', for Stack Release
- Factor DFi for Ground. Level or Split Wake Release mrem
- mrem / Ci' mrem yruCi yruCi sec m3 Kr-83m 3.5E-9 1.1E 16 7.56 E-2 Kr-85m 1.2E-4 3.8E-12 1.17 E3 Kr-85 1.7E-6 5.5E-14 1.61 El Kr-87 5.1E-4 1.6E-11 5.92 E3 Kr-88 1.4E-3 4.4E 11 1.47 E4 Kr 89 6.6E-4 2.1E-11 1.66 E4 Kr-90 -- -- 1.56 E4 Xe-131m 3.1E-5 9.7E 13 9.15 El Xe 133m 2.3E-5 7.3E-13 2.51 E2 Xe-133 2.5E-5 8.0E-13 2.94 E2 Xe 135m 2.5E-4 7.8E-12 3.12 E3 Xe-135 1.9E-4 6.0E-12 1.81 E3 Xe-137 5.4E-5 1.7E 12 1.42 E3 Xe 138 8.0E-4 2.5E-11 8.83 E3 Xe-139 1.6E-5 5.2E-13 5.02 E3 Ar-41 9.7E-4 3.1E-11 8.84 E3
' Based on reference meteorology; applicable at the site boundary,1,250 meters NNW of the ERP.
1 l
i l
\
Table 3-4 Transfer Factors for Maximum Dose To A Person Beyond The Site And Exclusion Area Boundary Due To t l
Radioactive Noble Gasea Radionuclide Dose Transfer Factors a
Aye,1 .
Ayy. Api 1
mrad mrad mrad pCi pCi sec/m i pCi sec/m 8 Kr-83m 2.6E 14 6.1E-7 9.13E-6 Kr-85m 4.0E 12 3.9E5 6.24E-5 Kr-85 5.8E-14 5.4E-7 6.18E-5 Kr-87 1.7E 11 2.0E-4 3.26E-4 Kr-88 4.6E 11 4.8E-4 9.28E-5 Kr-89 2.2E-11 5.5E-4 3.36E-4 Kr-90 -- 5.2E-4 2.48E-4 Xe 131m 1.1E 11 4.9E-6 3.52E-5 Xe 133m 8.7E-13 1.0E-5 4.69E-5 Xe-133 9.0E 13 1.1E-5 3.33E-5 Xe-135m 8.3E-12 1.1E-4 2.34E 5 Xe 135 6.3E 12 6.1E-5 7.79E 5 Xe-137 1.8E-12 4.8E-5 4.02E-4 Xe 138 2.7E-11 2.9E 4 1.51E 4 Ar 41 3.2E-11 2.9E-4 1.04E-4
' Dose at NNW site boundary 1
1 Table 3 5 l Transfer Factors for Maximum Dose To A Person Beyond Site and Exclusion Area Boundary Duc To Radioactive Noble Gases Radionuclide Dose Transfer Factors a,b PYes.1 Pyy. Spi 1
mrem mrem mrem 3
Ci pCi sec/m 8 pCi sec/m s Kr-83m 1.6E-16 2.4E-9 --
Kr-85m 2.4E 12 3.7E-5 4.6E-5 Kr 85 3.0E 14 5.1E-7 4.2E-5
\
Kr-87 7.9E 12 1.9E 4 3.1E-4 1 Kr-88 2.3E 11 4.7E-4 7.5E-5 Kr-89 6.7E-12 5.3E-4 3.2E-4 Kr-90 -- 4.9E-4 2.3E-4 1 Xe-131m 7.7E-13 2.9E-6 1.5E 5 Xe 133m 5.9E-13 8.0E-6 3.1E-5 Xe-133 6.9E 13 9.3E-6 9.7E-6 Xe-135m 3.3E 12 9.9E-5 2.3E 5 Xe-135 3.7E 12 5.7E-5 5.9E-5 Xe-137 5.1E-13 4.5E-5 3.9E 4 Xc 138 1.2E 11 2.8E 4 1.3E-4 Ar-41 1.5E-11 2.8E-4 8.5E 5 l
j
' Receptor located at 1.1 miles west of Station
' Based on reference meteorology at Cooper Station l
l l
l 1
f APPENDIX A (DELETED) l l
A1 1/27/89 l
l l
l APPENDIX B 1
l REFERENCE METEOROLOGICAL DATA l
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 l
used to generate tables of reference values of X/Q, depleted X/Q, and D/Q herein.
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. 0 0 0 0- 0- 0 0- 0- 0- 0- 0- 0- 0-
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NROS 8 8 8 8 9 8 8 8 9 9 9 9 9 8 8 8 E AI I
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0-E 1 E E E E E E E E 6 8 0 8 7 5 3 4 4 4 2 9 2 4 3 0 2 1 1 1 1 1 3 3 4 5 6 6 8 8 2 3 9 9 9 9 9 9 8 8 8 9 9 9 9 9 9 9 0 0 0- 0- 0- 0 0- 0- 0- 0- 0- 0- 0-5 E E E E E E E E E E E E E E E E 4 1 6 1 8 4 5 7 1 0 5 2 2 1 9 0 6 6 7 6 5 8 1 1 1 4 4 5 6 4 8 7 R
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l i 9 9 0 01 0 10 91 9 0 0 0 0 9 9 9 9
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4 E E E E E E E E E E E E E E E E 4 4 0 6 1 8 2 1 8 5 1 0 2 0 6 5 2 1 8 9 8 7 1 2 4 2 5 6 1 2 1 3 9 9 9 9 9 9 9 9 0 0 0 9 9 9 9 9
. 0- 0- 0- 0- 0- 0- 0- 0- 1 1 1 0- 0- 0- 0- 0-5 3 E E E E E E E E E E E E E E E E 2 6 6 8 3 0 7 6 2 9 4 1 1 8 8 1 3 2 1 1 1 1 1 3 8 3 7 1 2 2 5 6 9 9 9 9 9 9 9 9 9 9 9 9 9 9 8 8 0 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-S
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NS 8 9 9 9 9 9 9 8 9 9 9 9 8 8 8 8 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0-OI ID T 5 E
1 E E E E E E E E E E E E E E E T C 2 2 4 3 4 5 7 8 6 2 6 1 0 9 9 4 AD I 1 9 7 6 7 6 8 RR TANT R 1 5 5 3 6 1 1 2 2 NDOS I 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 8 ENITD 5 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0- 0 -
0- 0- 0- 0-CAAR )
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7 2
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6 6
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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.
l C-1
I 4
SAMPLE DESCRIPTION - TYPE LOCATION SAMPLE TYPES AND SAMPLE LOCATIONSW (See Sample Station Locations Map - Figures C-1 and C-2)
]
Sample Etalian Samole Descriotion - Tvoe and Location i No.1 Type: (1) Air Particulate and Charcoal Filters (2) Environmental Thermoluminescent Dosimetry j Location: Outside the northwest edge of fence, east of the gate to the LLRW storage pad on the CNS site, NW%, S32, T5N, RIGE, 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 approximately 275 feet west of the former Jefferson Broady farmstead, SWM, S32, T5N, RIGE, 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, SEM, S18, T5N, i RIGE, Nemaha County, Nebraska.
No. 4 Type: (1) Air Particulate and Charcoal Filters (2) Environmental Thermoluminescent Dosimetry Location: Located K 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 one fourth mile east of Langdon, ,
Missouri, on north side of road, west of railroad tracks, SW%, i 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," south side of road, SW corner of the intersection, NW%, S34, T64N, l R42W, Atchison County, Missouri. l C2 05/16/94
Sample Station Samole Descriotion - Tvoe and Location No.7 Type: (1) Air Particulate and Charcoal Filters (2) Environmental Thermoluminescent Dosimetry Location: 300 yards east of Highway 67 at Nemaha on north side of road, .
SW%, 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 Mark T. Moore 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 % mile north on west side of road, SW%, S26, T6N, R15E, Nemaha County, Nebraska.
No.10 Type: (1) Air Particulate and Charcoal Filters I (2) Environmental Thermoluminescent Dosimetry Location: One mile north of Barada, Nebraska, in SW corner of intersection, NEM, S14, T3N, RIGE, Richardson County, Nebraska.
No.11 Type: (1) Water - Ground Location: Plant well water supply header at well pits, NW%, S32, T5N, RIGE, Nemaha County, Nebraska.
No.12 Type: (1) Water - River Location: Sample (1) will be taken (monthly) from the Missouri River immediately upstream from the Plant Intake Structure (River Mile 532.5). :During periods when unsafe 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, approximately 20 yards east of county road, SE%, S30, T5N, R16E, Nemaha County, l Nebraska.
C-3 03/06/96
1 1'
Sample Station Samole Descrintion - Tyne and location No. 28 Type: (1) Water - River (2) Fish j (3) Sediment from Shoreline I (4) Food Products Broadleaf Vegetation Location: Samples (1), (3) and (4) are taken from the Missouri River or its shore, below the Plant Discharge Flume Outfall near River Mile 530. Sample (2) is taken from the Missouri river one-half to three miles downstream from the plant site.
No. 35 Type: (1) Fish (2) Water River (Alternate Site)
Location: Sample (1) will be taken twice a year from the Missouri River about one to three miles above intake structure. During periods when unsafe conditins warrant, Station 35 may be used as an alternate to Station 12 (upstream collection site) for sample type (2).
No. 44 Type: (1) Environmental Thermoluminescent Dosimetry Location: Two miles south of Auburn stop light, % mile south of Auburn Country Club on Highway #75, % mile east of Highway #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 south of Rulo, Nebraska (out of Main Header Flow Meter), SW%, S20, TIN, R18E, Richardson County, Nebraska.
No.5G Type: (1) Environmental Thermoluminescent Dosimetry Location: One and one-fourth mile south and west of Langdon, Missouri, on Highway "U", on the right side of the highway, Bill Gebheart farm, NWM, S23, T64N, R42W, Atchison County, Missouri.
No.58 Type: (1) Environmental Thermoluminescent Dosimetry i l
Location: Three miles south of Brownville, Nebraska, on county road, at the southeast corner of the intersection, with the farm road l leading to Sample Station No. 2, SEM, S31, T5N, R16E, Nemaha County, Nebraska.
No.59 Type: (1) Environmental Thermoluminescent Dosimetry Location: One mile SSE of the CNS Elevated Release Point,50 yards west of the levee at the south boundary of NPPD property, SE%, S32, T5N, R16E, Nemaha County, Nebraska.
C4 05/20/97
Sample Station Samole Descriotion - Type and Location No.61 Type: (1) Milk - Nearest Producer Location: 1 mile west of Brownville, NE, on Highway 136, then one mile north on on mile east county southroad, turn side of right road, and proceed Raymond Gentertapproxiately/4, dairy, NW 1 S13, T5N, RI5E, Nemaha County, Nebraska.
No. 66 Type: (1) Environmental Thermoluminescent Dosimetry Location: Two miles south of Nemaha Nebraska, on Highway 67 east side of highway. Mrs. I.ola kennedy farm. (NW%, Section 19, T4N, RIGE) Nemaha County, Nebraska.
No. 67 Type: (1) Environmental Thermoluminescent Dosimetry Location: Two miles west of Brownville, Nebraska, on U.S. Highway #136, then north 1% miles on county road, then east % mile, on north side of road, Walter Parkhurst farm, NE%, S11, T5N, R15E, Nemaha County, Nebraska.
No.71 Type: (1) Environmental Thermoluminescent Dosimetry Location: Two miles cast of Phelps City, Missouri, on U.S. Highway #136, then south 1% miles on county road, then west 'l mile, Tom Boatman farm, SE%, SG, T64N, R41W, Atchison County, Missouri.
No. 79 Type: (1) Environmental Thermoluminescent Dosimetry Location: 17/8 miles south of Brownville, Nebraska, on the cast side of the paved road. NPPD property. (SE%, Section 30, T5N, RIGE)
Nemaha County, Nebraska.
No. 80 Type: (1) Environmental Thermoluminescent Dosimetry Location: 2-1/8 miles south of Brownville, Nebraska, on the cast side of l the paved road. NPPD property. (NE%, Section 31, T5N, RIGE) l Nemaha County, Nebraska.
No. 81 Type: (1) Environmental Thermoluminescent Dosimetry Location: 2-3/8 miles south of Brownville, Nebraska', in the northeast aved county road and the CNS corner access of theNPPD road. intersection of the p(NE%, Section 31, T5N, RIGE) property.
l Nemaha County, Nebraska.
No. 82 Type: (1) Environmental Thermoluminescent Dosimetry Location: 7/8 mile south of Cooper Nuclear Station in a field. NPPD property, (SW%, Section 32, T5N, RIGE) Nemaha County, Nebraska.
No. 83 Type: (1) Environmental Thermoluminescent Dosimetry Location: 2% miles south of Nemaha, Nebraska, on Highway 67, then east one mile to the junction of the driveway and county road on the ,
east side of the driveway. Leroy Kennedy, (NE%, Section 19, l T4N, RIGE) Nemaha County, Nebraska. {
l 03/06/96 C-5
l l
i Sample q Station Samnle Descrintion - Tyne and Location No. 84 Type: (1) Environmental Thermoluminescent Dosimetry Location: 2% miles west of Brownville, Nebraska, on the south side of U.S.
Highway 136, west of Locust Grove School, Bruce L. Solie (NW%, Section 22, T5N, R15E) Nemaha County, Nebraska.
No. 85 Type: (1) Environmental Thermoluminescent Dosimetry Location: One mile east of Brownville, Nebraska, on U.S. Highway 136, then north % mile on the east side of the county road.
Scott Leseberg, (NE%, Section 33, T65N, R42W) Atchison County, Missouri. ,
No. 86 Type: (1) Environmental Thermoluminescent Dosimetry {
1 Location: One mile west of Phelps City, Missouri, on U.S. Highway 136, then north 1% miles on Highway "D" - on the west side of )
Highway "D". Mrs. Olin (Mildred) Harmes, (SE%, Section 22, T65N, R42W) Atchison County, Missouri.
{
No.87 Type: (1) Environmental Thermoluminescent Dosimetry l; i
Location: One mile west of Phelps City, Missouri, on U.S. Highway 136, l then south % mile on a county road and then 3/4 mile west on a l county road to the end of the road. Robert Graf, (SW%,
Section 3, T64N, R42W) Atchison County, Missouri. J No. 88 Type:(1) Environmental Thermoluminescent Dosimetry '
Location: One mile west of Phelps City, Missouri, on U.S. Highway 136, then south two miles at the end of the county road.
David MeyerKorth, (NW%, Section 11, T64N, R42W) Atchison County, Missouri.
No. 89 Type: (1) Environmental Thermoluminescent Dosimetry 1
Location: 2% miles south of Phelps City, Missouri, on Highway "U", then I
% mile west in the southeast corner of the county road intersection. Gertrude Rosenbohm, (NE%, Section 14, T64N, R42W) Atchison County, Missouri.
No. 90 Type: (1) Environmental Thermoluminescent Dosimetry Location: 1% miles west and 3/4 mile south of Langdon, Missouri, on Highway "U", then % mile west. Garth Green, (SW%,
Section 23, T64N, R42W) Atchison County, Missouri.
C.G 03/0G/96
Sample Station Samole Description Tvoe and Location No.91 Type: (1) Environmental Thermoluminescent Dosimetry Location: % mile west of Rock Port, Missouri, on the south side of the intersection of U.S. Highway 136 and U.S. Highway 275 at the water tower. Richard H. and Vicki Cook, (NW%, Section 28, TG5N, R41W) Atchison County, Missouri.
No. 94 Type: (1) Environmental Thermoluminescent Dosimetry Location: % mile south of Langdon, Missouri, on the west side of the road.
Max Peeler, (NE%, Section 24, T64N, R42W) Atchison County, Missouri.
No.90 Type:(1) Food Products - Broadleaf Vegetation Location: Approximately 1 mile south of Brownville, NE, along paved road in the road ditch in Sector R, SW %, S19, T5N, RIGE, Nemaha County, Nebraska.
No. 99 Type: (1) Milk (Nearest Producer)
Location: 1% miles south of Shubert, Nebraska, on the west side of Highway 67. James Zentner dairy. (NE%, S24, T3N, R15E),
Richardson County, Nebraska.
No.100 Type: (1) Milk (Other Producer)
Location: Two miles south and one mile west of Shubert, Nebraska, Dick James dairy. (SW%, S23, T3N, R15E), Richardson County, Nebraska.
No.101 Type:
} NOTES:
'd Sample Station numbers missing from sequence are inactive or discontinued Sample Stations.
1 C-7 05/20/97
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APPENEIX D CDAM SPECIFICATIONS
l OFFSITE DOSE ASSESSMENT MANUAL APPENDIX D TABLE OF CONTENTS D 1.0 Use and Application . . . . . . . . ... . D 1.0-1 D 2.0 De fi ni ti ons . . . . . . . . . . ...... .. . D 2.0-1 0 3.0 ODAM Specifications Applicability . . . . . . . . . . . D 3.0-1 0 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 0 3.1.3 Liquid Effluents Dose . . . . . . . . . . . . . . . D 3.1-7 D 3.1.4 Outside Temporary Storage of Radioactive Liquid . D 3.1-10 0 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 0 3.2.3 Iodine and Particulates .. ....... D 3.2-5 0 3.2.4 Offgas Treatment System . . . . . . . . . . .. D 3.2-10 0 3.2.5 Exhaust Ventilation Treatment System . . . . . . . D 3.2-12 D 3.2.6 Hydrogen Concentration ........... . . D 3.2-14 0 3.2.7 Primary Containment Venting and Purging . . .. D 3.2-15 0 3.3 INSTRUMENTATION D 3.3.1 Liquid Effluent Monitoring .... . .. D 3 3-1 D 3.3.2 Gaseous Effluent Monitoring . . . . . . . ... D 3.3-7 D 3.4 LIQUID /GASE0US DOSE D 3.4.1 Liquid / Gaseous Effluents Dose . . . . . . . . . . D 3.4-1 D 3.5 SOLID RADI0 ACTIVE WASTE D 3.5.1 Solid Radioactive Waste . . . . . . . . . . . . . . D 3.5-1 D 4.0 HONITORING PROGRAM D 4.1 Monitorin Program Compliance . . . . . . . . . . . D 4.1-1 0 4.2 Monitorin Program Concentration . . . . . . . . . D 4.2-1 0 4.3 Monitorin Program Dose . . . . . . . . . . . D 4.3-1 0 5.0 MISCELLAN0US PROGRAMS / REPORTS D 5.1 Interlaborator Comparison Program . . . . . . . . D 5.1-1 D 5.2 Annual Radiol ical Environmental Re) ort .. . . D 5.2-1 0 5.3 Annual Radioac ive Effluent Release leport . . . . D 5.3-1 D 5.4 Special Reports . . . . . . . . . . . . . . . . . . D 5.4-1 D 5.5 Major Changes to Radioactive Waste Treatment Systems (L1 quid. Gaseous, and Solid) . ... . . .D 5.5-1 CNS ODAM iii REVISION 1
Use and Application D 1.0 1 0 1.0 USE AND APPLICATION The Offsite Dose Assessment Manual (0 DAM) Specifications are contained in Section 3.0 of this appendix. They contain operational recuirements.
Surveillance Requirements, and re rting requirements. Adcitionally. the Required Actions and associated C letion Times for degraded Conditions are specified. The format is consiste with the Technical Specifications (Appendix A to the CNS Operating License).
Die definitions contained in Technical Specifications Section 1.1.
" Definitions." apply to the ODAM Specifications. Defined terms are shown in all capital letters, consistent wi th the Technical Specifications.
The rules of usage for the 00AM 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."
I Figure D2.a-1 l
1 D 2.0 DEFINITIONS N. y
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Figure 02.a-1 (Page 1 of 1)
Site and Exclusion Area Boundaries CNS 00AM D 2.0-1 REVISION 0
00AM Specifications Applicability 0 3.0 0 3.0 00AM Specifications Applicability The 00AM Specifications are subject to Technical Specifications Section 3.0.
" Limiting Condition for Operation (LCO) Applicability and Surveillance Requirement (SR) Applicability." with the following exceptions:
- 1. LCO 3.0.6. regarding support / supported system ACTIONS is not applicable to 00AM Specifications.
- 2. LC0 3.0.7. regarding allowances to change specified Technical Specifications is not applicable to 00AM Specifications.
- 3. Section 3.0 requirements are not applicable when so stated in notes within individual specifications.
I I
l l
CNS 00AM D 3.0-1 REVISION 0 l
= _. _
Liquid Effluents Concentration 0 3.1.1 0 3.1 LIQUID EFFLUENTS D 3.1.1 Liquid Effluents Concentration l DLC0 3.1.1 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 i radionuclides other than dissolved or entrained noble l gases: and I
- b. 2 x 10" pC1/ml total activity concentration for dissolved or entrained noble gases.
APPLICABILITY: At all times.
ACTIONS l
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. l Exclusion Area Boundary due to radioactive licuid effluent exceecs limits.
l l
Liquid Effluents Concentration D 3.1.1 SURVEILLANCE RE0UIREMENTS 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 00AM to verify that the with ODAM average concentration beyond the Site and Section 2.4 Exclusion Area Boundary does not exceed DLC0 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 l
LIQUID SAMPLE LOWER l RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE LIMIT OF DETECTION l TYPE TYPE FREQUENCY FREQUENCY ANALYSIS (LLD)(h)
Batch Waste Grab Each batch (a) Each batch (a) Principal Gamma 5 X 104 pCi/ml(i)
Release Tanks (c) sample Emmitters (j)(k) 1131 1 X 10' pCi/ml Grab sampic One batch / 31 days (b) Dissolved and I X 10' pCi/mi I 31 days (a) Entrained Gases (gamma emitters)
Proportional Each batch (a) 31 days (b) 11-3 1 X 10' pCi/ml Composite of grab samples (f) Gross Alpha 1 X 10' pCi/ml 92 days (b) Sr-89 5 X 10' pCi/mi Sr-90 5 X 10* pCi/mi Fe-55 1 X 10' uCi/mi Plant Service Grab Sample 7 days 7 days (b) Principal Gamma 5 X 10' 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 10' pCi/ml(i)
Discharge (c) Composite of Emitters (j)(k)
Grab Samples (g) 1 131 1 X 10' pCi/mi Grab Sample 31 days 31 days (b) Dissolved and I X 10' pCi/mi Entrained Gases (samma 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 10' pCi/ml Composite of Grab Samples (g) Gress Alpha 1 X 10' pCi/mi 92 days (b) St-89 5 X 10* pCi/mi Sr-90 5 X 10* pCi/ml Fe-55 1 X 10*pci/ml (a) Complete prior to each release.
(b) Analysis may be performed after release.
(c) A beech release is the discharge ofliquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and then thoroughly miaed.
(d) A grab ==ple of plant service water emuent shall be analyzed at least once each week in accordance with Table D3.1.1-1. Plant 4
Service Water Emment in the event the radioactivity concentration in a sample eaceeds 3 X 10 pCi/ml. or in the event the plant service water emuent monitor indicates the presence of an activity concentration greater than 3 X 10' pCi/mt, sampling and analysis according to Table D3.1.1-1. Plant CW n Discharge, shall commence and shall be performed as long as the condition persists.
(c) A contmuous release is the discharge ofliquid wastes of a nondiscrete volume; e.g., from a volume of system that has an input flow during the continuous release.
i Table D3.1,1-1 (Page 2 of 2)
Radioactive Liquid Waste Sampling and Analysis 1
(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):
(4.66XS6 )
LLD= 1 (EXVX2.22)(r)e ~"' {
l Where:
LLD is the 'a priori" lower hmit 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 picocune, '
Y is the fractional radiochemical yield (when applicable),
A is the radioactive decay constant for the particular radionuclide, and at is the clapsed 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 l theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the l 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.
(1) 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 10#/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 value specified in 10 CFR 20, Appendix B Table 2, Column 2.
(j) The principal gamma emmitters for which the LLD specification will apply are exclusively the following radionuclides: Mn-54, j 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 l 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 104pc/ml and the sample is analyzed when the instrumentation is once again available.
l 1
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l Liquid Waste Concentration D 3.1.2 D 3.1 LIQUID EFFLUENTS D 3.1.2 Liquid Waste Concentration l DLC0 3.1.2 The concentration of radioactive materials in liquid wastes l from pre-release analysis shall be s .01 pCi/ml, excluding i
tritium and noble gases.
APPLICABILITY: At all times.
ACTIONS l
NOTES--------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LCO 3.0.4 is not applicable.
, CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A.1 Appropriate parts of Prior to liquid radioactive materials the liquid radwaste waste discharge in liquid wastes from treatment system l pre-release analysis shall be used to l
> .01 yC1/ml, reduce the !
excluding tritium and concentration.
noble gases.
1 (continued) l 1
l '1 i
l l 0 1
Liquid Waste Concentration D 3.1.2 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Prepare and submit a 31 days J associated Completion Special Report to the following the j Time not met. NRC pursuant to end of the j Specification D 5.4 quarter in which l AtlQ that identifies the limit was l equipment or exceeded Radioactive liquid subsystems not waste being discharged OPERABLE and the without treatment in reason for the excess of .01 Ci/ml, inoperability, excluding tritium and action (s) taken to noble gases. restore the inoperable equipment to OPERABLE status and a summary description of the action (s) taken to prevent a recurrence.
1 CNS 00AM D 3.1-6 REVISION 0
i i
Liquid Effluents Dose D 3.1.3 D 3.1 LIQUID EFFLUENTS D 3.1.3 Liquid Effluents Dose DLC0 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. s 1.5 mrem to the total body or s 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
NOTES------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose due to A.1 Pre are and submit a 31 days radioactive material S ial Report in following the in licuid effluents 1 eu of any other end of the beyonc the Site and re rt. pursuant to quarter in which Exclusion Area S cification D 5.4 the limit was Boundary exceeds the t the NRC which exceeded limit. identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken.
(continued)
Liquid Effluents Dose D 3.1.3 ACTIONS (coninued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose due to B.1 Prepare and submit a 31 days radioactive material Special Report. in in licuid effluents lieu of any other beyonc the Site and report, pursuant to Exclusion Area Specification D 5.4 Boundary exceeds two to the NRC which 1) times the limit. defines actions to be taken to reduce 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 Coo >er Station during tie calender year through the period covered by the calculation was s 75 mrem to the thyroid and s 25 mrem to the total body and all other body organs.
i
, SURVEILLANCE RE0UIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.3.1 Perform an assessment of compliance with 31 days DLCO 3.1.3.
(continued)
CNS 00AM D 3.1-8 REVISION 0
l l
l Liquid Effluents Dose D 3.1.3 SURVEILLANCE REWIREMENTS (continued) 1 SURVEILLANCE FREWENCY DSR 3.1.3.2 Project a prospect of compliance with In any quarter DLC0 3.1.3 for radioactive liquid releases in which without radwaste system in operation. Radioactive liquid releases are made and the radwaste :
system is not operated.
1 1
Outside Temporary Storage of Radioactive Liquid D 3.1.4 D 3.1 t,IQUID EFFLUENTS D 3.1.4 Outside Temporary Storage'of Radioactive Liquid DLC0 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
NOTES------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Level of radioactivity A.1 Suspend addition of Immediately exceeds the limits of radioactive material. 1 TS 5.5.8.b.
r A.2 Begin measures to Immediately reduce content to within the limits of TS 5.5.8.b.
MD Prior to A.3 Describe the events submittal of leading to the next Annual condition in the Radioactive Annual Radioactive Effluent Release l Materials Release Report Report.
l CNS ODAM D 3.1-10 REVISION 0 l
l
l l
Outside Temporary Storage of Radioactive Liquid D 3.1.4 i
SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY DSR 3.1.4.1 Sample and analyze radioactive liquid 7 days during located in unprotected outdoor temporary addition of licuid storage tanks for level of radioactive racioactivity, liquid to the tanks I
)
J i
I l
l l
CNS 00AM D 3.1-11 REVISION 0 1
I Gaseous Effluents Concentration D 3.2.1 :
)
i D 3.2 GASE0US EFFLUENTS l D 3.2.1 Gaseous Effluents Concentration DLC0 3.2.1 The dose rate beyond the Site and Exclusion Area Boundary (Figure 02.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, I-131. I-133, and radioactive material in particulate form with half lives 2 8 days, s 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 s 3 months and:
- 2. The dose rate due to other radionuclides is averaged over s 31 days.
APPLICABILITY: At all times.
ACTIONS
_______________________________.-----NOTES------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Dose rates beyond the A.1 Decrease release rate Immediately f
Site and Exclusion to comply with the Area Boundary due to limits, radioactive gaseous effluents exceeds limits.
l l CNS ODAM D 3.2-1 REVISION 0 1
I l
Gaseous Effluents Concentration D 3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
1 DSR 3.2.1.1 Perform an assessment of compliance for 31 days DLC0 3.2.1(b).
l l
)
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Noble Gases Dose D 3.2.2 D 3.2 GASE0US EFFLUENTS
( D 3.2.2 Noble Gases Dose DLC0 3.2.2 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, s 5 mrad during any calender quarter and s 10 mrad during any calender year; and
- b. For beta radiation, s 10 mrad during any calender quarter and s 20 mrad during any calender year.
APPLICABILITY: At all times.
1 ACTIONS
_____________________________________ NOTES------------------------------------
- 1. LC0 3.0.3 is not applicable
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated air dose A.1 Prepare and submit a 31 days due to radioactive Special Report followi the t noble gases ond the pursuant to end of t e Site and Exclu ion Specification D 5.4 quarter in which Area Boundary exceeds to the NRC in lieu of the limit was the limit. any other report exceeded which identifies the l 1 cause(s) and defines i l the corrective actions taken.
(continued)
I CNS 00AM 0.3.2-3 REVISION 0 )
Noble Gases Dose 1
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 l releases and prevent I recurrence and 2) l results of an exposure analysis including effluent -
pathways and direct radiation to G determine whether the dose or dose i commitment to a j Member of the Public l due to radiation and l radioactive releases from Coo)er Station during t1e calender year through the period covered by the {
calculation was s 75 mrem to the thyroid and s 25 mrem to the total body or any other body organ.
i SURVEILLANCE RE00IREMENTS SURVEILLANCE FREQUENCY DSR 3.2.2.1 Perform an assessment of compliance for 31 days (
DLC0 3.2.2.
Iodine and Particulates j
D 3.2.3 D 3.2 GASEOUS EFFLUENTS 0 3.2.3 Iodine and Particulates DLC0 3.2.3 The dose to a Member of the Public due to I-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. s 7.5 mrem to any organ during any calender quarter:
and
- b. s 15 mrem to any organ during any calender year.
APPLICABILITY: At all times.
ACTIONS
.............._.....................-NOTES------------------------------------
- 1. LC0 3.0.3 is not applicable.
l
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose due to A.1 Prepare and submit a 31 days I-131 I-133 and Special Report, in following the radioactive material lieu of any other end of the i in particulate form report, pursuant to quarter in which having a half-life Specification D 5.4 the limit was
> 8 days beyond the to the NRC which exceeded.
Site and Exclusion identifies the Area Boundary exceeds cause(s) for the limit. exceeding the limit (s) and describes the corrective action taken.
(continued) i CNS ODAM D.3.2-5 REVISION 0 E--_ ._--_ _ o
(
Iodine and Particulates D 3.2.3 4
CONDITION REQUIRED ACTION COMPLETION TIME l B. Calculated dose due to B.1 Prepare and submit a 31 days )
I-131, 1-133 and Special Report, in l l radioactive material lieu of any other l l in particulate form report, pursuant to I l having a half-life Specification D 5.4 l
> 8 days beyond the to the NRC which 1) l Site and Exclusion defines actions to be
! Area Boundary exceeds taken to reduce two times the limit. releases and prevent l recurrence and 2) results of an exposure analysis including effluent pathways and direct radiation to !
determine whether the l l dose or dose commitment to a Member of the Public due to radiation and radioactive releases from Cooper Station was s 75 mrem to the
! thyroid and s 25 mrem to the total body or any other body organ.
l l
i SURVEILLANCE RE0UIREMENTS i SURVEILLANCE FREQUENCY 1
l l DSR 3.2.3.1 Perform radioactive gaseous waste sam) ling In accordance l and activity analysis on effluents otler with than noble gases. Table 03.2.3-1 l DSR 3.2.3.2 Perform a dose assessment to determine 31 days I
compliance with DLC0 3.2.3.
l 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 DE1EC110N(LLD)
TYPE TYPE FREQUENCY FREQUENCY ANALYSIS (h) d
- 1. Elevated Release Grab Sampic 31 days (c) 31 days (c) Principal Gamma 1 X 10 pCi/nd (i)
Point (ERP) Emitters (O 4
Grab Sampic 92 days (a) 92 days H-3 1 X 10 pCi/nd Charcoal Sample Connnuous (b) 7 days (d) 1-131 1 X 10-t2pCi/ml 1-133 1 X 10~ pCi/ml Particulate Sample Cononuous (b) 7 days (d) Principal Gamma i X 10'" pCi/ml (i)
Emiaers (0 (I 131, Others)
Composite Continuous (b) 92 days Sr-89 1 X 10'" pCi/ml Particulate Sampic (c)
Sr-90 1 X 10'" pCi/mi Gross Alpha 1 X 10'" pCi/ml Noble Gas Monitor Conunuous (b) Continuous Gross Noble 1 X 104 pCi/ml .
Gases (g) i (Beta, Gamma) l
- 2. Reactor Grab Sample 31 days (c) 31 days (c) Principal Gamma 1 X 10 pCi/nd (i)
Building Vent Emitters (O 4
Grab Sample 92 drys (a) 92 days H-3 1 X 10 pCi/mi Charcoal Sampic Conunuous (b) 7 days (d) 1-131 1 X 10-i2pCi/ml 1133 1 X 10- pCi/ml Particulate Sampic C=i-=>s (b) 7 days (d) Principal Gamma I X 10'" pCi/ml(i)
Emiacrs (0 (I-131, Others)
Composite Contmucus (b) 92 days Sr-89 1 X 10-" pCi/mi Particulate Sample (c)
Sr90 1 X 10-" pCi/ad Gross Alpha 1 X 10'" pCi/ml Noble Gas Monitor C"-=>! (b) C"--a Gross Noble I X 104pCi/mi Gases (g)
(Beta, Gamma)
- 3. Augmenced Grab Sample 31 days (c) 31 days (c) Principal Gamma i X 10 pCi/ml(i)
Radwasee Emiteers (0 Building Vent 1 X 104 pCi/ad l
Grab Sample 92 days (a) 92 days H-3 Charcoal Sample t'-i-=n (b) 7 days (d) 1-131 1 X 10'" pCi/ml (continued)
1 Iodine and Particulates D 3.2.3 i
Table D3.2.3-1 (Page 2 of 3) l Radioactive Gaseous Waste Sampling and Analysis l l
GASEOUS SAMPLE LOWER RELEASE SAMPLE SAMPLE ANALYSIS SAMPLE LIMIT OF TYPE TYPE FREQUENCY FREQUENCY ANALYSIS DETECTION (LLD)(h)
- 3. (continued) Charcoal Sample Comuasous (b) 7 days (d) 1-133 1 X 104 ' pCi/mi Particulate Sample Od-=2! (b) 7 days (d) Principal Gamma 1 X 10'" pCi/mi (i)
Emitters (0 (I-131, Others)
Composite Continuous (b) 92 days Sr-89 1 X 10'" pCi/ml Particulate Sample (c)
St-90 1 X 10'" pCi/mi Gross Alpha 1 X 10'" pCi/ml Noble Gas Monitor Cd=== (b) Contuanous Gross Noble ! X 104pCi/mi Gases (g)
(Beta, Gamma) 4
- 4. Turbine Building Grab Sample 31 days (c) 31 days (c) Principal Gamma 1 X 10 pCi/ml(i)
Vent (Gaseous) Eminers (f) 4 Grab Sample 92 days (a) 92 days H-3 1 X 10 pCi/ml Charcoal Sample Contmuous (b) 7 days (d) 1-131 1 X 10'" pCi/mi 1133 1 X 104 ' pCi/mi Particulate Sample Cw==>s (b) 7 days (d) Principal Gamma i X 10'" pCi/ml(i)
Eminers (f)
(I-131. Others)
Composite Contmuous (b) 92 days Sr-89 1 X 10'" pCi/mi Particulate Sample (c)
St-90 1 X 10'" pCi/ml Gross Alpha 1 X 10'" pCi/ml Noble Gas Monitor c"--as (b) C"--as Gross Noble ! X 104 pCi/ml Gases (g)
(Beta, m-)
(a) A H-3 grab sample wHl also be taken when abe renesor vessel head is renzved. This sample will be taken at the ERP or Reactor Boinding Vent whichever will be representative dependent upon the head renoval vacuum procedure (b) 7he ratio of the ===ple flow raec a) the sampled stream flow rate shall be known for the time period covered by each dose or dose rate e=h1=h made in ="- with Specifications D 3.2.1, D 3.2.2 and D 3.2.3.
(c) Analyses shall also be performed following an increase as indicated by the gaseous release monnor of greater than 50% in the steady state release, aher factoring out increases due to power changes or other operational A-.w which could alter the mixture of rad == Wides (d) Analysis shad also be perfonned following an hacrease as i=daad by the gaseous release momsor of greater than 50% in the sacady state release, aher factoring out increases due to power changes or other opersoonal occurrences, which could aher the miJuure of rad ==d** When sangnes callacead 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 correspondans LLD's may be increased by a factor of 10.
(continued)
Iodine and Particulates D 3.2.3 Table D3.2.3-1 (Page 3 of 3)
Radioactive Gaseous Waste Sampling and Analysis (c) A quarterly composite particulate sampic shall include a portion of each week's particulate samples conected during the quarter.
(f) The principal samma emiucts for which the IID specification will apply are esclusively the fonowing 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, Co40, Zn45, Mo-99, Cs-134. Cs-137, Cc-141, and Cc-144 for particulate enussaons. 'this list does not mean that only these nuclides are to be detected and reported. Other peaks which are nacasurable and identifiable, together with the above nuclides, shall also be identified and reported. Nuclides which are below the 11D for the analyses should not be reported as being present at the 11D level for that nuclide. When unusual circumstances cause 11D's higher than required for more than 31 days, the reasons shah be documented in the Annual Radioactive Efiluent Release Report.
(g) The nobic gas contmuous monitor shah 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 samgde that will be detected with 95% probability (5% probabuity of falsely concludmg that a blank observation represents a *real" signal.)
For a particular men =ement system (which may include radiochemical separation):
(4.66Xs )
(E)(V)(2.22) (Y[
Where:
11D is the "a priori
- lower limit of detectaon as described above (as picoeurie per unit mass or volume),
s,is the starriard deviation of the background counting rate or of the counung rate of a blank sarcple as appropriate (as counts per minute),
E is the counung 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 picoeurie, Y is the fractional radiochemical yield (when applicable),
l is the radioactive decay constant for the particular radionuclide, and at is the elapsed time between midpoint of sample collection and time of countmg (for plant effluents, not environmental samples).
The value of s used in the calculation of the 11D for a detection system shall be based on the actual observed vanance of the Wm " counting rate or of the countag rate of the blank samples (as appropnate) rather than on an unverified theoreticaDy predicted variance. In calculating the I1D for a radL=rtnie determined by gamma-ray .r om. guy, the background shall
. include the typical contributaons of other radaonuclides normany present in the sample:.. Typical vahaes of E, V, Y and At shad l
be used in the calculation.
(1) For certain radionuclades with low ganuna yield or low energies, or for certain radbinaclide =hanes, it may not be possible to men =e radaotmaclides in concentrations near the 11D. Under these circumstances, the 11D may be increased inversely rord,mwl to the magnitude of the gamma yield (i.a., I x 10*/I, where I is the photon abundance expressed as a decimal fraction), but in no case shall the 11D, 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.
I Offgas Treatment System D 3.2.4 l
t 1
! D 3.2 GASEOUS EFFLUENTS I D 3.2.4 Offgas Treatment System DLC0 3.2.4 Gaseous releases discharged through the Offgas Treatment System shall have at least one train of charcoal adsorbers in service.
APPLICABILITY: Main condenser air ejector in service, except during startup or shutdown with reactor < 10% rated power or when system l l cannot function due to low offgas flow.
1 ACTIONS l -------------------------------------NOTES------------------------------------
- 1. LCO 3.0.3 is not applicable.
- 2. LCO 3.0.4 is not applicable.
l CONDITION REQUIRED ACTION COMPLETION TIME A. Gaseous releases A.1 Restore release of 7 days discharged without gaseous discharge via either train of charcoal adsorbers.
charcoal adsorbers in service.
B. Required Action and B.1 Prepare and submit a 31 days associated Com Special Report following the Time not met. pletion pursuant to end of tfie Specification D 5.4 quarter in which t'othe NRC which the release l identifies the occurred.
inoperable equixnent l and describes tie l corrective action taken.
l CNS 00AM D 3.2-10 REVISION 0 1
L
Offgas Treatment System D 3.2.4 SURVEIILANCE REOUIREMENIS SURVEILLANCE FREQUENCY DSR 3.2.4.1 Verify o)eration of the Offgas Treatment In accordance System clarcoal 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 DLC0 3.2.5. Condition B. when radioactive material in gaseous effluent is released without treatment.
I Exhaust Ventilation Treatment Systems (EVTS)
D 3.2.5 I
D 3.2 GASE0US EFFLUENTS D 3.2.5 Exhaust Ventilation Treatment Systems (EVTS)
DLC0 3.2.5 The Exhaust Ventilation Treatment Systems (EVTS) shall be )
operated to treat radioactive materials in effluent air. 1 APPLICABILITY: When radioactive material in gaseous effluent is being released via the associated pathway.
ACTIONS
NOTES------------------------------------
- 1. LCO 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME I
A. Radioactive material in A.1 Ensure SR 3.2.5.1 is 31 days gaseous effluent met.
released without treatment.
B. Air is discharged B.1 Prepare and submit a 31 days without treatment for Special Report following the
> 31 days, pursuant to end of the Specification D 5.4 quarter in which ABEl to the NRC which the release identifies the occurred.
The projected dose to a inoperable equi ment Member of the Public and describes tie due to activity in air corrective action effluent via that taken.
pathway exceeds 0.3 mrem to any body organ.
l CNS 00AM D 3.2-12 REVISION 0 l
l I
Exhaust Ventilation Treatment System D 3.2.5 SURVEILLANCE REOUIREMENTS SURVEILLANCE FREQUENCY l
DSR 3.2.5.1 Project the prospect of compliance with Every 31 days DLC0 3.2.5. when radioactive material in gaseous effluent is released without treatment.
l l
Hydrogen Concentration D 3.2.6 D 3.2 GASE0US EFFLUENTS D 3.2.6 Hydrogen Concentration DLC0 3.2.6 The concentration of hydrogen in the augmented offgas treatment system downstream of the recombiners shall be limited to s 2% by volume.
APPLICABILITY: During augmented offgas treatment system operation.
ACTIONS
NOTES------------------------------------
- 1. LCO 3.0.3 is not applicable.
)
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME 1 l
1 A. Concentration of A.1 Restore concentration 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hydrogen exceeds to within limits.
limits.
CNS 00AM 0 3.2-14 REVISION 0
Primary Containment Venting and Purging D 3.2.7 0 3.2 GASE0US RELEASES' D 3.2.7 Primary Containment Venting and Purging DLCO 3.2.7 Venting and purgin of the primary containment shall be through the Standb Gas Treatment System.
- - - -This
- 1. - - -specification
- - - - - - - - -does
- - - not
- - -apply
- - - -to- Normal
- NOTEVenti
- 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
.... ..............................--NOTE-------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LCO 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Requirement not met A.1 Suspend all venting Immediately and purging of the primary containment.
1 CNS ODAM D 3.2-15 REVISION 0
Primary Containment Venting and Purging D 3.2.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ;
1 DSR 3.2.7.1 The primary containment shall be determined Once within 4 to be aligned for venting or purging hours prior to through the Standby Gas Treatment System. venting or purging of the primary containment 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 containment.
l CNS ODAM D 3.2-16 REVISION 0
Liquid Effluent Monitoring 0 3.3.1 D 3.3 INSTRUMENTATION D 3.3.1 Liquid Effluent Monitoring DLC0 3.3.1 The liguid effluent radiation monitoring instrumentation channels shown on Table 03.3.1-1 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 DLC0 3.1.1 are not exceeded.
APPLICABILITY: According to Table D3.3.1-1.
ACTIONS
NOTES------------------------------------
- 1. LCO 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
- 3. Separate condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A. Liquid effluent A.1 Suspend liquid Immediately radiation monitoring effluent radiation instrumentation release monitored by channel alarm and trip the inoperable setpoint less channel.
conservative than required.
08 A.2 Declare channel Immediately !
(continued)
CNS 00AM D 3.3-1 REVISION 0
t Liquid Effluent Monitoring i D 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME i B. One or more channels B.1 Enter the Condition Immediately l inoperable. referenced in Table D3.3.1-1 for I the channel.
l B.2.1 Restore ino erable 31 days I
l channel (s) o OPERABLE status.
08 B.2.2 In lieu of any other In accordance report, explain in with the Annual j the Annual Radioactive Radioactive Effluent Effluent Release l Release Report why Report the instrument was frequency.
not repaired in a .
l timely manner. !
C. As required by C.1 Analyze a minimum of Prior to Recuired Action B.1 2 independent sam)1es initiating a anc referenced in in accordance wit 1 release Table D3.3.1-1. Table D3.1.1-1.
MD l C.2 --------NOTE--------- .
Determination Action I and Verification Action will be performed by two separate technically .
qualified members of . l the Facility Staff. !
)
Determine and Prior to independently verify initiating a the release rate release calculations and discharge valving. ,
l (continued)
Liquid Effluent Monitoring D 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME D. As required by D.1 Collect and analyze a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Recuired Action B.1 grab sample for gross anc referenced in beta radioactivity or E Table D3.3.1-1. gross gamma radioactivity (as Once per applicable) at a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> lower limit of thereafter detection s 10 6 pCi/ml.
E. As required by E.1 Estimate flow rate 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Recuired Action B.1 during actual anc referenced in release. M Table D3.3.1-1.
Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> I thereafter F. Required Action and F.1 Suspend liquid Immediately i associated Completion effluent releases Time for Condition C monitored by the or E not met. inoperable channel (s).
G. Required Action and G.1 Initiate a Problem Immediately associated Completion Identification Report Time for Condition D for investigation of not met. the compliance failure.
i i
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 day on which indication of flow during periods of continuous, release. Seriodic. or
]atch 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 ,
l l
l 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.
(continued)
CNS 00AM D 3.3-4 REVISION 0
Liquid Effluent Monitoring D 3.3.1 SURVEILLANCE REQUIREMENTS (continued)
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 00AM D 3.3-5 REVISION 0
Liquid Effluent Monitoring D 3.3.1 Table D3.3.1-1 Radioactive Liquid Effluent Monitoring Instrumentation APPLICABILITY CONDITION OR On!ER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITIONS OPERABLE ACTION B.! REQUIREMENTS
- 1. Gross Beta or Gamma Radioactivity Monitors Providing Automatic isolation
- a. Liquid Radwaste (a) Ig 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 Line DSR 3.3.1.4 DSR 3.3.1.5 DSR 3.3.1.7
- 3. Flow Rate Measurement Devices
- a. Liquid Radwaste (a) 1 E DSR 3.3.1.2 Effluent Line DSR 3.3.1.5 DSR 3.3.1.8 (a) During releases via this pathway.
(b) Set so alarm and automatically close the waste discharge valve prior to exceeding the limits of Dif0 3.1.1.
CNS 00AM D 3.3-6 REVISION 0
Gastous Effluznt Monitoring D 3.3.2 l
D 3.3 INSTRUMENTATION 1 D 3.3.2 Gaseous Effluent Monitoring DLCO 3.3.2 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.
APPLICABILITY: According to Table D3.3.2-1.
ACTIONS NOTES
- 1. LCO 3.0.3 is not applicable.
- 2. LCO 3.0.4 is not applicable.
- 3. 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 instrumentation monitored by inoperable channel alarm and trip . channel.
setpoint less conservative than QB required. Immediately l
A.2 Declare channel inoperable.
(continued)
l.
Gastous Effluent Monitoring D 3.3.2 i
CONDITION REQUIRED ACTION COMPLETION TIME 1
l B. One or more channels B.1 Enter the Condition immediately inoperable, referenced in Table D3.3.2-1 for the l channel.. '
AND 1
- B.2.1 Restore inoperable 31 days l channel (s) to OPERABLE status.
B.2.2 In lieu of any other report, in accordance with explain in the Annual the Annual Radioactive Effluent Radioactive Effluent l Release Report why the Release Report j instrument was not frequency.
l repaired in a timely manner.
I (continued) l i
1 1
l l CNS ODAM D 3.3-8 REVISION 0
i Gaseous Efflu:nt Monitoring D 3.3.2 l REQUIRED ACTION COMPLETION TIME CONDITION C.1 Ensure the offgas delay immediately C. As required by Required Action B.1 system is not bypassed.
and referenced in Table D3.3.2-1.
6NQ Immediately C.2 Ensure the Elevated Re: case Point Monitoring j
nobh gas activity monitor is OPERABLE.
AND 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> C.3 Restore inoperable ,
channels to OPERABLE status.
Required Action and D.1 Be in MODE 2 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> D.
associated Completion Time for Condition C not met. l 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 1
(continued) 1 1
I l
i Gaseous Efflu:nt Monitoring D 3.3.2 l CONDITION REQUIRED ACTION COMPLETION TIME l
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 I and referenced in Table D3.3.2-1. AND l l
Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ;
thereaner l 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 l sampling completion l
G. As required by G.1.1 Verify one Function 2.a immediately l Required Action B.1 monitor OPERABLE and referenced in Table D3.3.2-1. AND j G.1.2 Monitor recombiner immediately exhaust temperature M
G.2.1.1 Collect gas sample 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
l AND AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter G.2.1.2 Analyze gas sample 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 AND G.2.2.1 Verify one Function 2.a monitor OPERABLE Immediately M
G.2.2.2 Monitor recombiner exhaust temperature immediately I
l (continued)
Gassous Efflurnt Monitoring D 3.3.2 CONDITION REQUIRED ACTION COMPLETION TIME H. Required Action and H.1 Discontinue operation of Immediately associated Completion the augmented offgas Time for Condition G treatment system.
not met.
- 1. ' 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.
QB 1.2.1 If auxiliary sampling Immediately equipment cannot be established within the specified completion time, initiate a Problem Identification Report to evaluate particulate and iodine effluent releases.
AND 1.2.2 Report this event in the in accordance with Annual Radioactive the Annual
) Effluent Release Report. Radioactive Effluent l Release Report i
l frequency.
J. Required Action and J.1 Discontinue effluent immediately associated Completion releases via this pathway.
Time for Condition E or F not met.
K. Function 1.a trip K.1 Close the offgas isolation immediately capability not valve maintained 6NQ 6ND K.2 initiate reactor shutdown immediately Radiation level exceeds 1.0 ci/sec (prior to 30 min. delay 6MR i line) for > 15 consecutive minutes 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 /> CNS ODAM D 3.3-11 02/09/99
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 (continued)
I CNS ODAM D 3.3-12 REVISION 0
Gaseous Effluent Monitoring D 3.3.2 SURVEILLANCE FREQUENCY DSR 3.3.2.8 Perform CHANNEL FUNCTIONAL TESl. 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.
DSR 3.3.2.10 Perform CHANNEL CALIBRATION. For Function 18 months 1.a. the time delay setting for closure of the steam jet air ejector isolation valves shall be s 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 i CHANNEL FUNCTIONAL TEST shall also I 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 l mode.
DSR 3.3.2.12 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months i
CNS 00AM D 3.3-13 REVISION 0
I Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 1 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation I
APP!JCABI!J1Y CONDmON OR ODIER MINIMUM REFERENCED SPECIAL CHANNE!Ji FROM SURVEILLANCE INTTRUMENT CONDmONS OPERABLE ACTION B.I REQUIREMENTS
- a. Noble Oas Activity (a) I g,) C DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.8 i DSR 3.3.2.10 l DSR 3.3.2.11 DSR 3.3.2.12 l b. Emuent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 l DSR 3.3.2.10 1
- 2. Augmented Offgas Treatment System Explosive Oas Monitoring System
- s. Hydrogen Monieor (c) 2 O DSR 3.3.2.1
( 25 monieor) DSR 3.3.2.4 DSR 3.3.2.6
- 3. Reactor Building Ventilation Monieorms System
- a. Noble Gas Activity (b) 1 F DSR 3.3.2.1 l Monisor DSR 3.3.2.3 l DSR 3.3.2.9 DSR 3.3.2.10
- b. lodine Sampler (b) I 1 DSR 3.3.2.2 Cartridge j
l c. Particulate Sampler (b) 1 1 DSR 3.3.2.2 Filter
- d. Emuent System Flow (b) 1 E DSR 3.3.2.1 Raee Measuring Device DSR 3.3.2.7 DSR 3.3.2.10
- e. tap Flow Raes (b) 1 E DSR 3.3.2.1 Meamring Device DSR 3.3.2.7 DSR 3.3.2.10
- f. Isolanon Monitor (d) (d) (d) DSR 3.3.2.5 DSR 3.3.2.11
(. . -4 (a) During operation of die steam jet air ejector (b) During seleases via this padeway (c) During mW offgas treatment system operation (d) See To Mical Specification 3.3.6.2 (e) Second t ennel must either be OPERABLE or be in the tripped condition.
l CNS 00AM D 3.3-14 REVISION 0 1
i Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 2 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDmON OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITION OPERABLE ACI10N B.I REQUIREMENTS
- 4. Elevated Release Point i Monitoring System
- a. Noble Gas Activity (b) 1 F DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10
- b. Iodine Sampler (b) 1 I DSR 3.3.2.2 Cartridge
- c. Partiutate Sampler (b) 1 I DSR 3.3.2.2 Filter
- d. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10
- c. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 Measuring Device DSR 3.3.2.7 DSR 3.3.2.10
- 5. Radwaste Building Ventilation Monitoring System
- a. Noble Gas Activity (b) 1 F DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10
- b. Iodine Sampler (b) 1 I DSR 3.3.2.2 Cartridge
- c. Particulate Sampler (b) 1 I DSR 3.3.2.2 Filter
- d. Effluent System Flow (b) 1 E DSR 3.3.2.1 Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10
- c. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 Measuring Device DSR 3.3.2.7 DSR 3.3.2.10
- 6. 'thrbene Building Ventilation Monitoring System
- a. Noble Gas Activity (b) 1 P DSR 3.3.2.1 Monitor DSR 3.3.2.3 DSR 3.3.2.9 DSR 3.3.2.10
- b. Iodine Sampler (b) 1 I DSR 3.3.2.2 Cartridge feontinued)
(b) Luring releases via this pathway CNS ODAM D 3.3-15 REVISION 0
3 Gaseous Effluent Monitoring D 3.3.2 Table D3.3.2-1 (Page 3 of 3)
Radioactive Gaseous Effluent Monitoring Instrumentation APPLICABILITY CONDmON OR OTHER MINIMUM REFERENCED SPECIAL CHANNELS FROM SURVEILLANCE INSTRUMENT CONDITION OPERABLE ACTION B.1 REQUIREMENTS l 6. (continued)
- c. Particulate Sampler (b) 1 I DSR 3.3.2.2 Filter
- d. Emuent System Flow (b) 1 E DSR 3.3.2.1 l
Rate Measuring Device DSR 3.3.2.7 DSR 3.3.2.10 l
- c. Sampler Flow Rate (b) 1 E DSR 3.3.2.1 l Measuring Device DSR 3.3.2.7 DSR 3.3.2.10 (b) During releases via this pathway l
Liquid / Gaseous Effluents Dose D 3.4.1 0 3.4 LIQUID /GASE0US DOSE D 3.4.1 Liquid / Gaseous Effluents Dose DLC0 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 s 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
__...__..........__......__.........-NOTES------------------------------------
- 1. LCO 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Estimated dose or dose A.1 Verify the condition Immediately commitment due to resulting in doses radiation and exceeding these radioactive releases limits is corrected.
exceeds limits.
(continued)
y Liquid / Gaseous Effluents Dose D 3.4.1 CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 -----------NOTE----------
associated Com This is the Special Time not met. pletion Report required by D 3.1.2. D 3.1.3. or D 3.2.3 supplemented with the following.
Submit a Special 31 days Report pursuant to
! Specification D 5.4.
including information specified in 40 CFR Part 190.11(b). This i
submission shall be deemed a timely request for variance in accord with 3rovisions of 40 CFR 3 art 190. The variance is granted until NRC staff action on the item is complete.
l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 3.4.1.1 Perform a cumulative dose calculation due 12 months to radioactive material in gaseous and liquid effluents to determine compliance with DLC0 3.4.1.
Solid Radioactive Waste D 3.5.1 D 3.5 SOLID RADI0 ACTIVE WASTE j D 3.5.1 Solid Radioactive Waste DLC0 3.5.1 The anroariate equipment of the solid radwaste system shall be OP HAB.E to process radioactive waste containing liquid andliquiddestinedfordisposalsubjectto10CFRPart61 i I
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
NOTES------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Container of waste A.1 Suspend delivery to Immediately does not comply with a carrier for 10 CFR Part 61.56. transport.
CNS 00AM D 3.5-1 REVISION 0
Solid Radioactive Waste D 3.5.1 SURVEILLANCE RE0UIREMENTS SURVEILLANCE FREQUENCY DSR 3.5.1.1 Sample and analyze the dewatered Prior to radioactive waste for pH. solidification l
of every 10th batch of dewatered l waste.
DSR 3.5.1.2 Inspect solidified or dewatered radioactive Prior to i waste to insure that there is no free capping each
! standing liquid on top of the solid waste. drum or High Integrity l
Container (HIC) l
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DSR 3.5.1.3 Record the following information for In accordance radioactive solid waste shipped offsite with j during the report period for the Annual 10 CFR 50.36a Radioactive Effluent Release Report per ;
the Reporting Requirements in Technical Specification 5.6.3:
a) Container burial volume, b) Total curie quantity (determined by l
measurement or estimate),
c) Principal gamma radionuclides (determined by measurement or estimate).
l d) Type of waste, and l
e) Solidification agent.
l CNS ODAM D 3.5-2 REVISION 0 E _ ___
Monitoring Program Compliance D 4.1 D 4.0 MONITORING PROGRAM D 4.1 Monitoring Program Compliance DLC0 4.1 The radiological environmental monitoring program shall be conducted as s:ecified in Table D4.1-1, using analytical techniques suc1 that the detection capabilities in {
Table 04.1-2 are achieved.
APPLICABILITY: At all times.
' ACTIONS
_________________________________----NOTES------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Radiological A.1 Prepare and submit May 15th environmental to the NRC in the following the monitoring program not Annual Radiological end of the year .
conducted as specified Environmental Report l in Table D4.1-1. the reasons for not conducting the program in accordance with Table 04.1-1 and the plans for preventing recurrence.
B. Environmental sampling B.1 Report in the Annual May 15th medium is not Radiological following the available from a Environmental Report end of the year sampling location as the cause and specified in location where Table D4.1-1. replacement sam were obtained. ples l
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1 Monitoring Program Compliance j
! D 4.1 l
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 4.1.1 Perform radiological environmental sampling In accordance ,
and analysis, with l Table D4.1-1 l 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 I
that yields' edible leafy vegetables, the l location of the nearest milk animal, and l the location of the nearest resident in each of the 16 meteorological sectors l within 3 miles of the Station.
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l DSR 4.1.3 Summarize results of radiological May 15th l 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 l CNS 00AM D 4.1-2 REVISION 0
Monitoring Program Compliance D 4.1 Table 04.1-1 (Page 1 of 2)
Radiological Environmental Monitoring Program EXPOSURE PATHWAY NUMBER OF SAMPLING AND TYPE AND FREQUENCY SAMPLE STATIONS COLLECTION FREQUENCY OF ANALYSIS AND/OR SAMPLE
- 1. Airborne Contmuous operation of sampler Radiciodine canister: Analyze
- a. Radiciodine and At least 5 locations with sample collection as required at least once per 7 days for Particulate by dust loading but at least once 1131.
per 7 days.
Particulate sample: Analyze for gross beta radioactivity a 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following fdter change. Performgamma isotopic (" analysis on each sample in which gross beta activity is >10 times the yearly mean of control samples.
Perform gamma isotopic (" !
analysis on composite (by location) sample at least once per 92 days.
At least 32 locations 'lhermoluminescent Dosimeters Gamma dose: At least once per
- 2. Direct Radiation (TLD)" exchange and read-out at 92 days.
least once per 92 days.
- 3. Waterborne At least 2 locations Collect a one (1) gallon grab Gamma isotopic (" analysis of
- a. River Water sample at least once per 31 days, each sample. Composite grab sample for tritium analysis at least once per 92 days.
Gamma isotopic" I and tritium
- b. Ground Water At least 2 locations Collect a one (1) gallon grab sample at least once per 92 days. analysis of each sample.
- c. Sedunent from At least I location Two (2) times a year, once in the Gamma isotopic (" analysis of Shoreline spring and osce in the fall. each sample.
- 4. Ingestion At least i location At least once per 15 days during Gamma isotopic (* and I-131
- a. Milk (nearest Peak Pasture Period"; at least once analysis of each sample.
Producer) per 31 days at other times.
- b. Mdk (other producers) At least 2 locations At least once per 92 days. Gamma isotopic (* and I-131 l
analysis of each sample.
(
- c. Fish Atleast 2 locations Two times per year (once in the Gamma isotopic (* analysis on summer and once in the fall). edible portions.
Ateetnpt to include the following:
- 1. Bonom feeding species
/
- 2. Middle-Top feeding species (consmued)
I CNS ODAM D 4.1-3 REVISION 0 L
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Monitoring Program Compliance D 4.1 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 (8and1131 kinds of broad leaf analysis vegetation grown nearest each of two different offsite locations of highest predicted anmaal average ground-level D/Q if milk sampling is not performed.
One sample of each of the Monthly when avaliable. Gamma isotopicidand 1-131 l similar broad leaf analysis vegetation grown 15-30 km distant in the least i prevalent wind direction if milk sampling is not performed-(a) Ge (Li) gamma isotopic analysis refers to high resolution Ge (Li) 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 countmg time, detector efficiency, samma energy, geometry, and detector background appropriate to the particular sample in question. ;
'Ihe following nineteen (19) nuclides shall be anclyzed routinely:
Bala-140 Cs-137 Ra226 Be-7 Fe 59 Ru-103 Ce 141 1-131 Ru-106 Ce 144 K-40 1h-228 Co-58 Nb 95 Zn-65 Co-60 Mn-54 Zr-95 Cs-134 Any radionuclide detected i.e., having a measured concentration greater than the LID, whether or not k is one of the 19 nuclides listed above, shan be regarded as present in the sample.
(b) Ther=*=ni-wat Dosuneters (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 I through September 30 of each year.
CNS 00AM 0 4.1-4 REVISION 0
T Monitoring Program Compliance D 4.1 Table D4.1-2 (Page 1 of 2)
Detection Capabilities for Environmental Sample Analysis LOWER LIMIT OF DETECTION"(LLD)"
AIRBORNE PARTICULATE FOOD WATER OR GAS FISH MILK PRODUCTS SEDIMENT (pCi/I) (pCi/m*) (pCi/ks, we0 (pCi/l) (pCi/kg, wet) (pci/kg, dry)
ANALYSIS gross beta 4 1 x 10'#
H-3 2000 Mn-54 15 130 Fe-59 30 260 Co-58 15 130 Co-60 15 130 Zn45 30 260 Zr-95 30 Nb-95 15 I-131 9" 7 x 10 2 1 60 Cs-134 15 5 x 10'8 130 15 60 150 4 18 80 180 Cs-137 18 6 x 10 150 Ba-140 60 60 La-140 15 15 (a) *lhis list does not mean that only these maclides are so be h and reported Other peaks which are measurable and identirable.
together wkh the above =elulas, shall also be identifud and reported.
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Monitoring Program Compliance D 4,1 Table 04.1-2 (Page 2 of 2)
Detection Capabilities for Environmental Sample Analysis I (b) The LLD is the *a priori
- smallest concentration of redmactive 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):
(4.66X: p LLD =
(EXVX2.22) (Y)(e )
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 approprime (as counts per minute),
E is the counting efficiency (as counts per transfonnation),
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 at is the elapsed time between san $le collection (or midpoint of the sampic collection period) and tinw of countag The value of s used in the calculation of the LLD for a detection system shall be based on the actual observed variance of the background countag raec or of the countmg 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 odwr radionuclides normauy 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 condnions rW-ily background fluctuations, unavoidably small sample sizes, the presence of interfering maclides, or other uncontrollable cira=*-a may render these LID's unachievable. la such cases, the contributing factors will be identified and descrihi in the Annual Radiological Envirosunental Report.
(c) LLD for drmking water.
[
Monitoring Program Concentration D 4.2 0 4.0 MONITORING PROGRAM D 4.2 Monitoring Program Concentration DLC0 4.2 Radioactivity concentrations in sampled medium from the radiological environmental monitorincI ram shall not
! exceed values specified in Table 05.11pr en averaged over a calender quarter.
l-APPLICABILITY: At all times.
(
ACTIONS
.....................................N0TES------------------------------------
- 1. LC0 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
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CONDITION REQUIRED ACTION COMPLETION TIME A. Radioactivity A.1 Prepare and submit 31 days concentrations of to the NRC a Special following the sampled medium from Report in accordance end of the the radiological with Specification quarter environmental D 5.4 which includes monitoring program an evaluation of any exceeds values release conditions.
specified in environmental factors Table D5.4-1. averaged or other conditions over a calender which caused the quarter which is value(s) to be attributable to exceeded.
release (s) from the Station.
(continued) l CNS 00AM D 4.2-1 REVISION 0
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Monitoring Program Concentation l
D 4.2 ACTIONS- (continued)
CONDITION REQUIRED ACTION COMPLETION TIME B. Radioactivity B.1 Report and explain in May 15th following the
! concentrations of the Annual sampled medium from Radiological end of the year the radiological Environmental Re) ort environmental the results of t1e monitoring program sample (s).
! exceeds values i specified in Table D5.4-1, averaged over a calender i
quarter which is not i attributable to release (s) from the
- Station.
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Monitoring Program Dose D 4.3 0 4.0 MONITORING PROGRAM D 4.3 Monitoring Program Dose DLC0 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
NOTES------------------------------------
- 1. LCO 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. ' Location (s) identified A.1 --------NOTE---------
at which the Only applicable if 1 calculated dose samples are associated with the reasonably attainable exposure pathway (s) at the new location.
exceeds 120% of the ---------------------
calculated dose at the maximum dose location Add new sampling At the next SRAB associated with like location (s) meeting pathways at a location identified having where sampling is maximum exposure conducted as specified potential to the in Table 04.1-1. radiological environmental monitoring program and Table D4.1-1.
AND A.2 Describe change made May 15th to Table D4.1-1 in following the l the Annual end of the year Radiological Environmental Report.
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Interlaboratory Comparison Program l 0 5.1 l !
D 5.0 MISCELLANE0US PROGRAMS / REPORTS D 5.1 Interlaboratory Comparison Program DLC0 5.1 Analyses shall be performed on radioactive materials supplied as part of the Interlaboratroy Comparison Program which has been approved by the NRC.
APPLICABILITY: At all times.
l ACTIONS l ........__
__________.....______.__--NOTES------------------------------------
l 1. LC0 3.0.3 is not applicable.
- 2. LC0 3.0.4 is not applicable.
CONDITION REQUIRED ACTION COMPLETION TIME A. Analyses not A.1 Report to the NRC in May 15th performed. the Annual following the Radiological end of the year Environuental Report the corrective
- actions taken to prevent recurrence.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DSR 5.1.1 Submit a brief summary of the results May 15th obtained from the Interlaboratory following the Comparison Program in the Annual end of the year l Radiological Environmental Report to Technical Specification 5'.6.2. pursuant 1
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i CNS ODAM D 5.1-1 REVISION 0 I
Annual Radiological Environmental Report D 5.2 0 5.0 MISCELLANE0US PROGRAMS / REPORTS D 5.2 Annual Radiological Environmental Report
}
The Annual Radiological Environmental Report shall include the following:
l 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 l 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 i DSR 4.1.2.
- c. Summarized and tabulated results in the format of Table D5.2-1 of analysis of samples recuired by the radiological environmental I monitoring program. anc 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 partici .
required by 0 5.1.pation in the Interlaboratory Comparison Program 1
CNS 00AM D 5.2-1 REVISION 0
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Annual Radioactive Effluent Release Report D 5.3 D 5.0 MISCELLANE0US PROGRAMS / REPORTS !
D 5.3 Annual Radioactive Effluent Release Report The Annual 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 DLC0 3.1.1.
DLC0 3.1.3, DLC0 3.2.1, DLC0 3.2.2 or DLC0 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.
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Special Re) orts
) 5.4 l
l D 5.0 MISCELLANEOUS PROGRAMS / REPORTS l D 5.4 Special Reports l
Special re) orts shall be submitted to the Director. Nuclear Reactor Regulation. l USNRC, Was11ngton. 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.
Wftcial 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) + Concentrat- on (2) + .. 2 1.0 Limit Level (1) Limit Leve' (2)
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 DLC0 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 CNS 00AM D 5.4-1 REVISION 1
l Special Reports D 5.4 !
1 Table 05.4-1 Reporting Levels for Radioactivity Concentrations in Environmental Samples Reporting Levels AIRBORNE PAR 11CULATE OR FISH FOOD PRODUCTS ANALYSIS WATER (pCill) GASES (pCi/m') (pCi/Kg, Wet) MILK (pCi/1) (pci/Kg, Wet)
H-3 2E + 4(a) 3B + 4(c)
M&M IE+3 3E + 4 Fe-59 4E + 2 1E+4 i Co-58 IE+3 3E + 4 f Co-60 3E + 2 1E+4 Zn45 3E + 2 2E + 4 Zr-Nb-95 4E + 2(b) 1-131 2 0.9 3 IE+2 Cs-134 30 10 IE+3 60 IE+3 Cs-137 50 20 2E + 3 70 2E + 3 Ba-La-140 2E + 2(b) 3E + 2(b)
(a) For drmking waaer samples. *lhis is the 40 CFR 141 value.
(b) Concentration of parent or daughter.
(c) For sampics of water not used as a source of drinking water.
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i Major Changes to Radioactive Waste Treatment Systems D 5.5 0 5.0 MISCELLANEOUS PROGRAMS / REPORTS D 5.5 Major Changes to Radioactive Waste Treatment Systems (Liquid, Gaseous, and Solid) l 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 I liquid effluents and in solid waste packaged for offsite shipment recuired to meet the DLC0's set forth in Specifications D 3.1.1, D 3.1.2 D 3.1.3, D E.1.4, i' 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).
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CNS ODAM D 5.5-1 REVISION 1 I l I J
y-0FFSITE DOSE ASSESSMENT MANUAL APPENDIX D TABLE OF CONTENTS - BASES i
B 3.1 LIQUID EFFLUENTS B 3.1.1 L1 uid Effluents Concentration . . . . . . . . . . B 3.1-1 i' B 3.1.2 Li uid Waste Concentration . . . . . . . . . . . . B 3.1-2 B 3.1.3 Li uid Effluents Dose . . . . . . . . . . . . . . . B 3.1-5 B 3.1.4 Ou side Temporary Storage of Radioactive Liquid . . B 3.1-9 8 3.2 GASE0US 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 8 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 j B 3.3 INSTRUMENTATION B 3.3.1 Liquid Effluent Monitoring . . . . . . . . . . . . B 3.3-1 i B 3.3.2 Gaseous Effluent Monitoring . . . . . . . . . . . . B 3.3-2 B 3.4 LIQUID /GASE0US EFFLUENTS DOSE B 3.4.1 Liquid / Gaseous Effluents Dose . . . . . . . . . . B 3.4-1 B 3.5 SOLID RADI0 ACTIVE WASTE B 3.5.1 Solid Radioactive Waste . . . . . . . . . . . . . . B 3.5-1 B 4.0 MONITORING PROGRAM B 4.1 Monitori Program Compliance . . . . . . . . . . . B 4.1-1 (
B 4.2 Monitorin Program Concentration . . . . . . . . . .B 4.2-1 B 4.3 Monitorin Program Dose . . . . . . . . . . . . . . B 4.3-1 ;
B 5.0 MISCELLANE0US PROGRAMS / REPORTS B 5.1 Interlaborator Comparison Program . . . . . . . . B 5.1-1 B 5.2 Annual Radiol ical Environmental Report. . . . . . B 5.2-1 B 5.3 Annual Radiol ical Effluent Release Report . . . . B 5.3-1 B 5.4 Special Reports . . . . . . . . . . . . . . . . . . B 5.4-1 8 5.5 Major Changes to Radioactive Waste Treatment Systems (L1guid, Gaseous, and Solid). . . . . . . . B 5.5-1 4 . . s. ,,.._.._s, . , . . . . . . _ ,,.,.___,,.____..,,,._.,..._.s
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s - - - + ~ . . 3 ,
i CNS 00AM iv REVISION 1 4
n 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)(1) 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 l release, routine sampling and monitoring for radioactivity is precautionary. An activity concentration of 3 x 10 6 Ci/mi in Service Water effluent is diluted in I the discharge canal to about 1.5% of the 10 CFR 20 Appendix B Table 2 Column 2 l 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.
l By monitoring Service Water continuously and liquid radwaste continuously during discharge with the monitor set to alarm or trip before the limit s)ecified in 10 CFR 20.1302 is exceeded, reasonable assurance of compliance witi Specification D 3.1.1 is provided. Verification that radioactivity in liquid effluent averaged i
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)(1) 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 com)liance. 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 I 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.
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Reporting by Special Reports and other reports required by the ODAM and Section 5.6 j l of Technical Specifications is used in lieu of reporting per 10 CFR 50.73. !
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l Liquid Waste Concentration B 3.1.2 {
B 3.1 LIQUID EFFLUENTS B 3.1.2 Liquid Waste Concentration BASES S)ecification D 3.1.2 implements the recuirements of 10 CFR Part 50.36a(a)(1) tlat operating procedures be establishec 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, A)pendix I,Section II be used when specified provides reasonable assurance tlat releases of radioactive material in liquid effluent will be kept as low as is reasonably achievable. The activity concentration. 0.01 gC1/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 2 total iodines 3.65 curies total others (less H )
3 IL1_
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 IL5fi total 2.5 manrem Therefore, population doses are about 0.5 manrem per curie of iodine 3 released and about 0.8 manrem per curie of other radionuclides (less H )
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)
CNS 00AM B 3.1-2 REVISION 0
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Liquid Waste Concentration !
B 3.1.2 BASES (continued) l The annual costs to operate the radwaste processing equi ment, neglecting credit for capital recovery, are estimated according to legulatory Guide 1.110 i to be: l Dirty Waste Ionex $ 88.000/yr Evaporator $114.000/yr l
Unit volume operating costs are about:
Cost to ion exchanger = $ 88.000 - $0.05/ gal 1.8E+6 gal Cost to evaporate = $114.000 - $0.10/ gal 1.2E+6 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 is not cost-beneficial is 1
6
$88.000 1 curie 10 Ci 1 manrem C- x x x i manrem curie $1,000 l 1.8E+6 gal x 3785 g l C - 0.013 gC1/ml (2) the activity concentration below which ' evaporation is not I costbeneficial is l $114.000 1 curie 106 pCi 1 manrem C x x x manrem curie $1,000 1.2E+6 gal x 3785 g C = 0.025 pCi/ml ,
Therefore, to one significant digit, radwaste treatment of liquids containing less than 0.01 Ci/ml is not justified.
(continued)
CNS 00AM B 3.1-3 REVISION 0 1
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Liquid Waste Concentration B 3.1.2 BASES (continued)
IDemonstration of Compliance with 10 CFR 50 Appendix I. Revision 1 and Supplement 2, Nebraska Public Power District, Cooper Nuclear Station. January 9,1978.
l CNS ODAM B 3.1-4 REVISION 0 L. -
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l Liquid Effluents Dose l
< B 3.1.3 B 3.1 LIQUID EFFLUENTS l B 3.1.3 Liquid Effluents Dose BASES l
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 3rocedural i details of RETS to the Offsite Dose Calculation Manual or to tie Process i Control Program.
Generic Letter 89-01 provides the guidance and justification for relocation of these " technical specifications" to the Offsite Dose Assessment (00AM) 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. A)pendix I.Section IV. These specifications state ODAM LIMITING CONDITIO45 FOR OPERATION (DLC0) 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 c liance with the Specifications. Action statements in these Specifications i lement l the requirements of 10 CFR Part 50.36(c)(2) and 10 CFR Part 50. Append x 1.
Section IV.A in the event an LC0 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 i
elsewhere in the Technical S>ecifications (are not an immediate safety concern) but do obligate NPP) to take the applicable Required Action in Specifications D 3.1.3. D 3.2.2 and D 3.2.3 (continued) l l
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CNS 00AM B 3.1-5 REVISION 0
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i Liquid Effluents Dose B 3.1.3 BASES (continued) 10 CFR Part 50 contains two distinctly separate statements of requirements l
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 l 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 Commicsion and the requirement are clearly stated in the Opinion of the Commission: relevant paragraphs from that 1
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 Ap>endix I that we adopt provides more specific guidance to licensees in t11s 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 Ccmmission within 30 days of the end of the calendar quarter.
2 The conclusion of the NRC Staff in the Appendix I Pulemaking Hearing agrees l
' 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) i CNS 00AM B 3.1-6 REVISION 0 l i
Liquid Effluents Dose B 3.1.3 BASES (continued) l The action to be taken by a licensee in the event a limiting condition is l exceeded, is stated in Appendix I.Section IV.A and in the opinion of the l Commission. 3 ODAM S l DSR 3.1.3.1, 3.1.3.2 3.2.2.1.
pecifications D and 3.2.3.1 3.1.3. D 3.2.2.
3.2.3.2 D 3.2.3 Station for Cooper and Surveillances conform i to this requirement. I Guidance for developing limiting conditions for o>eration for surveillance and i monitoring is included in Appendix I,Section IV.3. l 1
l Although "it is ex>ected that the annual releases of radioactive material in effluents from lig1t water cooled nuclear power reactors can generally be maintained within the levels set forth as numerical guides for design objectives in Section II" (Appendix 1.Section IV), 4no recommendation was made by either the Staff in its Concluding Statement or.by the Commission in its Opinion 5 that design ob:iective values should appear as. technical specification limits. T1e Opinion of the Commission and the statement of Appendix I are clear. Limiting conditions of operation (LC0) 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 LC0 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 l 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 I drinking water.
1N RC Commissioners, "031nion of the Commission," in the Appendix I Rulemaking hearing Docket Rm SC'2, p.101-102, April 30,1975, l 2 NRC Staff. " Concluding Statement of the Regulatory Staff." in the Appendix I Rule- making Hearing, Docket RM 502, pp.17, 69, 73,115. February, 1974.
3N RC Commissioners, p. 101
'NRC Staff, op. cit.
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NRC Commissioners, op. cit.
(continued) l CNS ODAM B 3.1-7 REVISION 0
Liquid Effluents Dose B 3.1.3 BASES (continued)
' 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 3rocedural details of RETS to the Offsite Dose Calculation Manual or to t1e Process Control Program.
CNS 00AM B 3.1-8 REVISION 0
l Outside Temporary Storage of Radioactive Liquid B 3.1.4 B 3.1 LIQUID EFFLUENTS l
B 3.1.4 Outside Temporary Storace of Radioactive Liquid l
BASES Custom Technical Specifications Bases did not exist. I CNS 00AM B 3.1-9 REVISION 0
Gaseous Effluents Concentration B 3.2.1 B 3.2 GASE0US EFFLUENTS B 3.2.1 Gaseous Effluents Concentration BASES DLC0 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 DLC0 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 s 500 mrem / year to the total body or to s 3000 mrem / year to the skin.
Radiciodines and radionuclides in particulate form are sampled with integrating samplers that permit assessment of the average release rate during each sample collection aeriod. By complying with DLC0 3.2.2 and 3.2.3 the average concentration aeyond 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. ;
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CNS 00AM B 3.2-1 REVISION 0 I
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Noble Gases Dose B 3.2.2
'B 3.2 GASE0US 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 00AM 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 aeriod of record used in the design objective Appendix I evaluation) atmospleric 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.
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CNS 00AM B 3.2-2 REVISION 0 j i
Iodine and Particulates l B 3.2.3 8 3.2 GASE0US EFFLUENTS .i B 3.2.3 Iodine and Particulates )
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BASES 1 l
This bases for Specification D 3.2.3 and Surveillances DSR 3.2.3.1 and 3.2.3.2 i are discussed in the bases for Specification D 3.1.3 and Surveillances OSR l 3.1.3.1 and 3.1.3.2.
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CNS 00AM B 3.2-3 REVISION 0
l Offgas Treatment System B 3.2.4 8 3.2 GASE0US 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 IID of Appendix I to 10 CFR i Part 50. The specified limits governing the use of a>propriate portions of ,
this system are specified as a suitable fraction of tie dose design objectives set forth in Sections II.B and II.C of Appendix I, 10 CFR Part 50, for gaseous effluents.
CNS 00AM B 3.2-4 REVISION 0 1
I Exhaust Ventilation Treatment Systems B 3.2.5 B 3.2 GASE0US 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 j
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 ap3ropriate portions of these systems be used when specified provides reasonaale 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 IID 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 GASE0US EFFLUENTS l 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 00AM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
Contrinm:nt B 3.2.7 j B 3.2 GASEOUS RELEASES B 3.2.7 Primary Containment Venting and Purging BASES This specification provides reasonable assurance that releases of lodine 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 lodine 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.
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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 00AM 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 0 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 00AM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
CNS 00AM B 3.3-1 REVISION 0 l
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o 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 00AM, 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 ex)losive gas mixtures in the augmented offgas treatment system. The OPERABI_ITY 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 provides an improved capability to of 1.0 ci/sec (prior to 30 min. delay)llow detect fuel pin cladding failures to a 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 adeguate 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 00AM and Section 5.6 of Technical Specifications is used in lieu of reporting per 10 CFR 50.73.
CNS 00AM B 3.3-2 REVISION 0
Liquid / Gaseous Effluents Dose B 3.4.1 B 3.4 LIQUID /GASE0US DOSE B 3.4.1 Liquid / Gaseous Effluents Dose BASES 1
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 I 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.
I CNS 00AM B 3.4-1 REVISION 0
Solid Radioactive Waste B 3.5.1
'B 3.5 SOLID RADI0 ACTIVE WASTE B 3.5.1 Solid Radioactive Waste BASES The OPERABILITY of the solid radwaste system ensures that the system will be l
available for use whenever solid radwastes require materials processing and packaging prior to being shipped offsite. This specification implements the l
requirements of 10 CFR Part 50.36a and General Design Criteria 60 of Appendix A to 10 CFR Part 50.
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Monitoring Program Compliance B 4.1 B 4.0 MONITORING PROGRAM l 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 sup)1ements the radiological effluent monitoring program by verifying tlat 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 04.1-1 is the minimum program which will be maintained. The Offsite Dose Assessment Manual (00AM) 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 for consumption by a child. To determine this minimum garden size the following assumptions were used. 1) that-20% of the garden was used for broad leaf vegetation (i.e.. similar to lettuce and cabbage), and 2) a growingionyieldof2kg/squaremeter.
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i CNS 00AM B 4.1-1 REVISION 0 l
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.
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i Monitoring Program Dose 8 4.3 '
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B 4.0 MONITORING PROGRAM B 4.3 Honitoring 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 04.1-1 at the time the new pathway (s) and locations are added.
I CNS 00AM B 4.3-1 REVISION 0 l
Interlaboratory Comparison Program B 5.1 B 5.0 MISCELLAN0US PROGRAMS / REPORTS B 5.1 Interlaboratory Comparison Program BASES The requirement for )articipation in a Interlaboratory Comparison Program is provided to ensure tlat 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. Participation in an Interlaboratory Com)arison Program is contingent u)on availability of samples supplied by the NRC or samples approved by the NRC.
A CNS 00AM B 5.1-1 REVISION 0
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l Annual Radiological Environmental Report )
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B 5.0 MISCELLAN005 PROGRAMS / REPORTS B 5.2 Annual Radiological Environmental Report i 1
BASES Custom Technical Specifications Bases did not exist.
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1 CNS 00AM B 5.2-1 REVISION 0
r Annual Radiological Effluent Release Report {
B 5.3 '
B 5.0 MISCELLAN0US PROGRAMS / REPORTS B 5.3 Annual Radiological Effluent Release Report l
BASES ]
Custom Technical Specifications Bases did not exist.
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CNS 00AM B 5.3-1 REVISION 0
Special Reports B 5.4 8 5.0 MISCELLAN0US PROGRAMS / REPORTS B 5.4 Special Reports BASES Custom Technical Specifications Bases did not exist.
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Major Changes to Radioactive Waste Treatment Systems B 5.5 B 5.0 MISCELLAN0US PROGRAMS / REPORTS B5.5Maj,orChangestoRadioactiveWasteTreatment' and.olid) Systems (Liquid, Gaseous, BASES Custom Technical Specifications Bases did not exist.
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