Text

Annual Radioactive Effluent Release Reports for 2010, Page 2-26 Through End
	ML111220088
Person / Time
Site:	Hatch, Vogtle, Farley
Issue date:	04/29/2011
From:	Ajluni M; Southern Nuclear Operating Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML111220082	List: ML111220085; NL-11-0811, Annual Radioactive Effluent Release Reports for 2010, Page 2-26 Through End; NL-11-0811, Edwin I. Hatch, Units 1 and 2, Joseph M. Farley, Units 1 and 2, and Vogtle, Units 1 and 2, Annual Radioactive Effluent Release Reports for 2010, Cover Letter Through Page 2-25;
References
	NL-11-0811
	Download: ML111220088 (270)
	v • d • e

FNP*ODCM

For streams that cannot be diverted or isolated, a new monitor setpoint must be established whenever: the radioactivity in the stream becomes detectable above the background levels of the applicable laboratory analyses; or the associated radioactivity monitor detects activity in the stream at levels above the established alarm setpoint.

When an elevated monitor setpoint is required for any of these effluent streams, it should be determined in the same manner as described in Section 2.3.2. However, special consideration must be given to Step 3. An allocation factor must be assigned to the normally low-radioactivity release pathway under consideration, and allocation factors for other release pathways discharging simultaneously must be adjusted downward (if necessary) to ensure that the sum of the allocation factors does not exceed 1. Sampling and analysis of the normally low-radioactivity streams must be sufficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.

2-26 Version 24 01/10

FNP*ODCM 2.4 LIQUID EFFLUENT DOSE CALCULATIONS The following sub-sections present the methods required for liquid effluent dose calculations, in deepening levels of detail. Applicable site-specific pathways and parameter values for the calculation of D~, A h , and CF iv are summarized in Table 2-5.

2.4.1 Calculation of Dose The dose limits for a MEMBER OF THE PUBLIC specified in Section 2.1.3 are on a per-unit basis. Therefore, the doses calculated in accordance with this section must be determined and recorded on a per-unit basis, including apportionment of releases shared between the two units.

For the purpose of implementing Section 2.1.3, the dose to the maximum exposed individual due to radionuclides identified in liquid effluents released from each unit to UNRESTRICTED AREAS will be calculated as follows (equation from Reference 1, page 15):

(2.12) where:

the cumulative dose commitment to the total body or to any organ t, in mrem, due to radioactivity in liquid effluents released during the total of the m time periods Atl .

the site-related adult ingestion dose commitment factor, for the total body or for any organ t, due to identified radionuclide i, in (mrem.mL)/(h * ~Ci).

Methods for the calculation of Ai~ are presented below in Section 2.4.2.

The values of Ai~ to be used in dose calculations for releases from the plant site are listed in Table 2-8.

the length of time period I, over which Cil and F/ are averaged for liquid releases, in hours.

Cil = the average concentration of radionuclide i in undiluted liquid effluent during time period I, in ~Ci/mL. Only radionuclides identified and detected above background in their respective samples should be included in the calculation.

the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA:

2-27 Version 24 01/10

FNP-ODCM (2.13) where:

ft = the average undiluted liquid waste f10wrate actually observed during the period of radioactivity release, in gpm.

Ft = the average dilution stream flowrate actually observed during the period of radioactivity release, in gpm.

Z = the applicable dilution factor for the receiving water body, in the near field of the discharge structure, during the period of radioactivity release, from Table 2-5.

Note: In equation (2.13), the product (F t x Z) is limited to 1000 cfs (= 448,000 gpm) or less. (Reference 1, Section 4.3.)

2.4.2 Calculation of Ait The site-related adult ingestion dose commitment factor, A it , is calculated as follows (equation adapted from Reference 1, page 16, by addition of the irrigated garden vegetation pathway):

(2.14) where:

5 1.14 x 10 = a units conversion factor, determined by:

106 pCilllCi x 10 3 mUL + 8760 h/y.

Uw = the adult drinking water consumption rate applicable to the plant site (Uy).

Ow = the dilution factor from the near field of the discharge structure for the plant site to the potable water intake location.

f...j = the decay constant for radionuclide i (h- 1). Values of f...j used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 15.

tw = the transit time from release to receptor for potable water consumption (h).

Ut = the adult rate of fish consumption applicable to the plant site (kg/y).

2-28 Version 24 01/10

FNP-ODCM SF; = the bioaccumulation factor for radionuclide i applicable to freshwater fish in the receiving water body for the plant site, in (pCi/kg)/(pCi/L) = (Ukg). For specific values applicable to the plant site, see Table 2-6.

tf = the transit time from release to receptor for fish consumption (h).

= the adult consumption rate for irrigated garden vegetation applicable to the plant site (kg/y).

= the concentration factor for radionuclide i in irrigated garden vegetation, as applicable to the vicinity of the plant site, in (pCilkg)/(pCi/L). Methods for calculation of CF iv are presented below in Section 2.4.3.

DF;,! = the dose conversion factor for radionuclide i for adults, in organ 't (mrem/pCi). For specific values, see Table 2-7.

2.4.3 Calculation of CF;v The concentration factor for radionuclide i in irrigated garden vegetation, CF iv in (Ukg), is calculated as follows:

For radionuclides other than tritium (equation adapted from Reference 3, equations A-a and A-9):

-M I r (1 -e-A.Eite ) I' B JI iv (1 -e-A.itb )] -).-th C'17 l'iv-* + e l (2.15)

[ Y;,AEi PAi

For tritium (equation adapted from Reference 3, equations A-9 and A-10):

(2.16) where:

M = the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage.

= the average irrigation rate during the growing season (L)/(m 2 .h).

r = the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation.

= the areal density (agricultural productivity) of leafy garden vegetation 2

(kg/m ).

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FNP-ODCM

= the fraction of the year that garden vegetation is irrigated.

= the crop to soil concentration factor applicable to radionuclide i. from Table 2-6 (pCi/kg garden vegetation)/(pCi/kg soil).

p 2

= the effective surface density of soil (kg/m ).

= the decay constant for radionuclide i (h -1). Values of Ai used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 15.

= the rate constant for removal of activity from plant leaves by weathering (h-\

1

= the effective removal rate for activity deposited on crop leaves (h- )

calculated as: AEi = Ai + Aw.

te = the period of leafy garden vegetation exposure during the growing season (h).

tb = the period of long-term buildup of activity in soil (h).

th = the time between harvest of garden vegetation and human consumption (h).

Lv = the water content of leafy garden vegetation edible parts (Ukg).

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FNP-ODCM Table 2-5 Parameters for Calculation of Doses Due to Liquid Effluent Releases Dose Calculation Receptor Locations:

Fish: Vicinity of plant discharge Drinking Water: None (Ref. 10)

Irrigated Garden Vegetation: Farms at River Mile 26 (Ref. 10)

Numerical Parameters:>

Parameter Reference 5 Ref. 2, Table A-1 OUy" Ref. 10 1.0 " Based on Ref. 1, Section 4.3.1 12 h" Ref. 3, Sec. A.2 21 kgly Ref. 3, Table E-5 24 h Ref. 3, Sec. A.2 64 kg/y Ref. 3, Table E-5 0.04 Ref. 16 0.126 U (m 2 h) Ref. 10, using pump capacity, garden size, and irrigation 10% of the time during growing season.

0.25 Ref. 3, Table E-15 2

2.0 kg/m Ref. 3, Table E-15 0.1 Ref. 10 240 kg/m 2 Ref. 3, Table E-15 1

0.0021 h- (Le., half-life =14 d) Ref. 3, Table E-15 1440 h (= 60 d) Ref. 3, Table E-15 5

1.31 X 10 h (= 15y) Ref. 3, Table E-15 24 h Ref. 3, Table E-15 0.92 Ukg Based on Ref. 11, Table 5.16 (for lettuce, cabbage, etc.)

Because there is no drinking water pathway downstream of the plant site, the consumption of drinking water is set to zero, and the default values of tw and Ow are used.

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FNP*ODCM Table 2*6 Element Transfer Factors Freshwater Fish Leafy Garden Element , Vegetation BF j Bjv+

H 9.0 E-01 4.8 E+OO C 4.6 E+03 5.5 E+OO Na 1.0 E+02 5.2 E-02 P 3.0 E+03 1.1 E+OO Cr 2.0 E+02 2.5 E-04 Mn 2.0 E+01 2.9 E-02 Fe 1.0 E+03 6.6 E-04 Co 1.0 E+02 9.4 E-03 Ni 1.0 E+02 1.9 E-02 Cu 1.5 E+02 1.2 E-01 Zn 1.0 E+02 4.0 E-01 Br 4.2 E+02 7.6 E-01 Rb 2.0 E+03 1.3 E-01 Sr 3.0 E+01 1.7 E-02 Y 2.5 E+01 2.6 E-03 Zr 2.0 E+02 1.7 E-04 Nb 1.0 E+02 9.4 E-03 Mo 1.0 E+02 1.2 E-01 Tc 1.5 E+01 2.5 E-01 Ru 1.0 E+01 5.0 E-02 Rh 1.0 E+01 1.3 E+01 Ag 2.3 E+OO 1.5 E-01 Sb 3.0 E+02 1.1 E-02 Te 2.0 E+03 1.3 E+OO I 2.0 E+01 2.0 E-02 Cs 2.0 E+02 1.0 E-02 Ba 4.0 E+01 5.0 E-03 La 2.5 E+01 2.5 E-03 Ce 2.0 E+02 2.5 E-03 Pr 2.5 E+01 2.5 E-03 Nd 2.5 E+01 2.4 E-03 W 1.2 E+03 1.8 E-02 Np 1.0 E+01 2.5 E-03 Bioaccumulation Factors for freshwater fish, in (pCi/kg)/(pCi/L). They are obtained from Reference 3 (Table A-1). except as follows: Reference 9 for P; Reference 2 (Table A-8) for Ag; Reference 8 for Mn, Fe, Co, Cu, Zn, Mo, Sb, Te, I, Cs, Ba, and Ce; and Reference 14 for Zr and Nb.

+ Crop to soil concentration factors, in (pCi/kg garden vegetation) per (pCi/kg SOil). They are obtained from Reference 3 (Table E-1), except as follows: Reference 2 (Table C-5) for Br and Sb.

2-32 Version 24 01/10

FNP-ODCM Table 2-7 Adult Ingestion Dose Factors

~1.0SE-07 I Nuclide Bone Liver T.Body Th- __ .,..l Kidney H-3 No Data 1.0SE-07 1.0SE-07 1.0SE-07 1.0SE-07 1.0SE-07 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 Cr-S1 No Data No Data 2.66E-09 1.S9E-09 S.B6E-10 3.S3E-09 6.69E-07 Mn-S4 No Data 4.S7E-06 B.72E-07 No Data 1.36E-06 No Data 1.40E-OS Mn-S6 No Data 1.1SE-07 2.04E-OB No Data 1.46E-07 No Data 3.67E-06 Fe-SS 2.7SE-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 1.09E-06 Fe-S9 4.34E-06 1.02E-OS 3.91 E-06 No Data No Data 2.BSE-06 3.40E-OS Co-SB No Data 7.4SE-07 1.67E-06 No Data No Data No Data 1.S1 E-OS Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-OS Ni-63 1.30E-04 9.01 E-06 4.36E-06 No Data No Data No Data 1.BBE-06 Ni-6S S.2BE-07 6.B6E-OB 3.13E-OB No Data No Data No Data 1.74E-06 Cu-64 No Data B.33E-OB 3.91E-OB No Data 2.10E-07 No Data 7.10E-06 Zn-6S 4.B4E-06 1.54E-OS 6.96E-06 No Data 1.03E-OS No Data 9.70E-06 Zn-69 1.03E-OB 1.97E-OB 1.37E-09 No Data 1.2BE-OB No Data 2.96E-09 Br-B3 No Data No Data 4.02E-OB No Data No Data No Data S.79E-OB Br-B4 No Data No Data S.21E-OB No Data No Data No Data 4.09E-13 Br-BS No Data No Data 2.14E-09 No Data No Data No Data No Data Rb-B6 No Data 2.11E-OS 9.B3E-06 No Data No Data No Data 4.16E-06 Rb-BB No Data 6.0SE-OB 3.21E-OB No Data No Data No Data B.36E-19 Rb-B9 No Data 4.01 E-OB 2.B2E-OB No Data No Data No Data 2.33E-21 Sr-B9 3.0BE-04 No Data B.B4E-06 No Data No Data No Data 4.94E-OS Sr-90 7.S8E-03 No Data 1.86E-03 No Data No Data No Data 2.19E-04 Sr-91 S.67E-06 No Data 2.29E-07 No Data No Data .~

~

All values are in (mrem/pCi ingested). They are obtained from Reference 3 (Table E-11),

except as follows: Reference 2 (Table A-3) for Rh-10S, Sb-124, and Sb-12S.

2-33 Version 24 01/10

FNP-ODCM Table 2-7 (contd) Adult Ingestion Dose Factors I~ Sr-92 Bone 2.1SE-06 Liver No Data T.Body 9.30E-08 Thyroid No Data Kidney No Data Lung No Data GI-LLI 4.26E-OS Y-90 9.62E-09 No Data 2.S8E-10 No Data No Data No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.S2E-12 No Data No Data No Data 2.67E-10 Y-91 1.41 E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-OS Y-92 8.4SE-10 No Data 2.47E-11 No Data No Data No Data 1.48E-OS Y-93 2.68E-09 No Data 7.40E-11 No Data No Data No Data 8.S0E-OS Zr-9S 3.04E-08 9.7SE-09 6.60E-09 No Data 1.S3E-08 No Data 3.09E-OS Zr-97 1.68E-09 3.39E-10 1.SSE-10 No Data S.12E-10 No Data 1.0SE-04 Nb-9S 6.22E-09 3.46E-09 1.86E-09 No Data 3.42E-09 No Data 2.10E-OS Mo-99 No Data 4.31E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2.47E-10 6.9SE-10 8.89E-09 No Data 1.06E-OS 3.42E-10 4.13E-07 Tc-101 2.S4E-10 3.66E-10 3.S9E-09 No Data 6.S9E-09 1.S7E-10 1.10E-21 Ru-103 1.8SE-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-OS Ru-10S 1.S4E-OS No Data 6.0SE-09 No Data 1.99E-07 No Data 9.42E-06 Ru-106 2.7SE-06 No Data 3.48E-07 No Data S.31E-06 No Data 1.7SE-04 Rh-10S 1.22E-07 S.S6E-OS S.83E-OS No Data 3.76E-07 No Data 1.41 E-OS Ag-110m 1.60E-07 1.4SE-07 8.79E-OS No Data 2.91E-07 No Data 6.04E-OS Sb-124 2.S1E-06 S.30E-OS 1.11 E-06 6.79E-09 No Data 2.18E-06 7.9SE-OS Sb-12S 2.23E-06 2.40E-OS 4.48E-07 1.9SE-09 No Data 2.33E-04 1.97E-OS Sb-126 1.1SE-06 2.34E-OS 4.1SE-07 7.0SE-09 No Data 7.0SE-07 9.40E-OS Te-12Sm 2.68E-06 9.71E-07 3.S9E-07 S.06E-07 1.09E-OS No Data 1.07E-OS Te-127m 6.77E-06 2.42E-06 8.2SE-07 1.73E-06 2.7SE-OS No Data 2.27E-OS Te-127 1.10E-07 3.9SE-OS 2.3SE-08 8.1SE-08 4.4SE-07 No Data 8.6SE-06 Te-129m 1.1SE-OS 4.29E-06 1.S2E-06 3.9SE-06 4.80E-OS No Data S.79E-OS Te-129 3.14E-OS 1.18E-OS 7.6SE-09 2.41E-OS 1.32E-07 No Data 2.37E-OS Te-131m 1.73E-06 S.46E-07 7.0SE-07 1.34E-06 8.S7E-06 No Data S.40E-OS Te-131 1.97E-OS S.23E-09 6.22E-09 1.62E-OS 8.63E-OS No Data 2.79E-09 2-34 Version 24 01/10

FNp*ODCM Table 2*7 (contd) Adult Ingestion Dose Factors li~ I

~

Q,.,n",

Thyroid Kidnev lung GI-lL!

Te-132 2.S2E-06 1.63E-06 1.S3E-06 1.BOE-06 1.S7E-OS No Data 7.71E-OS 1-130 7.S6E-07 2.23E-06 B.BOE-07 1.B9E-04 3.4BE-06 No Data 1.92E-06 1-131 4.16E-06 S.9SE-06 3.41E-06 1.9SE-03 1.02E-OS No Data 1.S7E-06 1-132 2.03E-07 S.43E-07 1.90E-07 1.90E-OS B.6SE-07 No Data 1.02E-07 1-133 1.42E-06 2.47E-06 7.S3E-07 3.63E-04 4.31E-06 No Data 2.22E-06 1-134 1.06E-07 2.BBE-07 1.03E-07 4.99E-06 4.SBE-07 No Data 2.S1 E-1 0 1-13S 4.43E-07 1.16E-06 4.2BE-07 7.6SE-OS 1.B6E-06 No Data 1.31E-06 Cs-134 6.22E-OS 1.4BE-04 1.21E-04 No Data 4.79E-OS 1.S9E-OS 2.S9E-06 Cs-136 6.S1E-06 2.S7E-OS 1.BSE-OS No Data 1.43E-OS 1.96E-06 2.92E-06 Cs-137 7.97E-OS 1.09E-04 7.14E-OS No Data 3.70E-OS 1.23E-OS 2.11E-06 Cs-13B S.S2E-OB 1.09E-07 S.40E-OB No Data B.01E-OB 7.91E-09 4.6SE-13 Ba-139 9.70E-OB 6.91 E-11 2.B4E-09 No Data 6.46E-11 3.92E-11 1.72E-07 Ba-140 2.03E-OS 2.SSE-OB 1.33E-06 No Data B.67E-09 1.46E-OB 4.1BE-OS Ba-141 4.71E-OB 3.S6E-11 1.S9E-09 No Data 3.31E-11 2.02E-11 2.22E-17 Ba-142 2.13E-OB 2.19E-11 1.34E-09 No Data 1.BSE-11 1.24E-11 3.00E-26 la-140 2.S0E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.2SE-OS la-142 1.2BE-10 S.B2E-11 1.4SE-11 No Data No Data No Data 4.2SE-07 Ce-141 9.36E-09 6.33E-09 7.1BE-10 No Data 2.94E-09 No Data 2.42E-OS Ce-143 1.6SE-09 1.22E-06 1.3SE-10 No Data S.37E-10 No Data 4.S6E-OS Ce-144 4.BBE-07 2.04E-07 2.62E-OB No Data 1.2'1 E-07 No Data 1.6SE-04 Pr-143 9.20E-09 3.69E-09 4.S6E-10 No Data 2.13E-09 No Data 4.03E-OS Pr-144 3.01 E-11 1.2SE-11 1.S3E-12 No Data 7.0SE-12 No Data 4.33E-1B Nd-147 6.29E-09 7.27E-09 4.3SE-10 No Data 4.2SE-09 No Data 3.49E-OS W-1B7 1.03E-07 B.61E-OB 3.01E-OB No Data No Data No Data 2.B2E-OS Np-239 1.19E-09 .4SE-11 No Data 3.6SE-10 No Data 2.40E~1 2-3S Version 24 01/10

FNP-ODCM Table 2*8 Site-Related Ingestion Dose Factors, Ail Nuclide Bone Liver T. Body Thyroid Kidney H-3 0.00 2.54E-01 2.54E-01 2.54E-01 2.54E-01 2.54E-01 2.54E-01 Na-24 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 1.34E+02 Cr-51 0.00 0.00 1.25E+00 7.45E-01 2.74E-01 1.65E+00 3.13E+02 Mn-54 0.00 2.28E+02 4.34E+01 0.00 6.77E+01 0.00 6.97E+02 Mn-56 0.00 8.69E-03 1.54E-03 0.00 1.10E-02 0.00 2.77E-01 Fe-55 6.58E+03 4.55E+03 1.06E+03 0.00 0.00 2.54E+03 2.61 E+03 Fe-59 1.02E+04 2.41E+04 9.22E+03 0.00 0.00 6.72E+03 8.02E+04 Co-58 0.00 1.78E+02 3.99E+02 0.00 0.00 0.00 3.61E+03 Co-60 0.00 5.17E+02 1.14E+03 0.00 0.00 0.00 9.71E+03 Ni-63 3. 14E+04 2.18E+03 1.05E+03 0.00 0.00 0.00 4.54E+02 Ni-65 1.72E-01 2.23E-02 1.02E-02 0.00 0.00 0.00 5.66E-01 Cu-64 0.00 8.07E+00 3.79E+00 0.00 2.04E+01 0.00 6.88E+02 Zn-65 1.17E+03 3.71E+03 1.68E+03 0.00 2.48E+03 0.00 2.34E+03 Zn-69 3.94E-08 7.54E-08 5.24E-09 0.00 4.90E-08 0.00 1.13E-08 Br-83 0.00 0.00 3.83E-02 0.00 0.00 0.00 5.52E-02 Br-84 0.00 0.00 1.22E-12 0.00 0.00 0.00 9.61E-18 Br-85 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Rb-86 0.00 9.74E+04 4.54E+04 0.00 0.00 0.00 1.92E+04 Rb-88 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Rb-89 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sr-89 2.23E+04 0.00 6.41E+02 0.00 0.00 0.00 3.58E+03 Sr-90 5.61E+05 0.00 1.38E+05 0.00 0.00 0.00 1.62E+04 Sr-91 7.07E+01 0.00 2.86E+00 0.00 0.00 0.00 3.37E+02 Sr-92 3. 33E-0 1 0.00 1.44E-02 0.00 0.00 0.00 6.60E+00 Y-90 4.47E-01 0.00 1.20E-02 0.00 0.00 0.00 4.74E+03 Y-91m 1.04E-11 0.00 4.01E-13 0.00 0.00 0.00 3.04E-11 Y-91 8.58E+00 0.00 2.30E-01 0.00 0.00 0.00 4.72E+03 Y-92 4.60E-04 0.00 1.35E-05 0.00 0.00 0.00 8.07E+00 Y-93 3.09E-02 0.00 8.54E-04 0.00 0.00 0.00 9.81E+02 Zr-95 1.45E+01 4.64E+00 3.14E+00 0.00 7.27E+00 0.00 1.47E+04 Zr-97 3.01E-01 6.07E-02 2.77E-02 0.00 9.16E-02 0.00 1.88E+04 Nb-95 1.47E+00 8.17E-01 4.39E-01 0.00 8.08E-01 0.00 4.96E+03 Mo-99 0.00 8.03E+02 1.53E+02 0.00 1.82E+03 0.00 1.86E+03 Tc-99m 5.60E-04 1.58E-03 2.02E-02 0.00 2.40E-02 7.76E-04 9.37E-01 All values are in (mrem*mL) 1 h*I.lCi). They are calculated using equation (2.14), and data from Table 2-5, Table 2-6, and Table 2-7. When "No Data" is shown for a radionuclide-organ combination in Table 2-7, AiT factors in this table are presented as zero.

2-36 Version 24 01/10

FNP*ODCM Table 2*8 (contd) Site-Related Ingestion Dose Factors, Ait Nuclide Bone Liver ~ Thyroid Kidney Lung ~

Tc-101 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Ru-103 4.6SE+00 0.00 2.00E+00 0.00 1.77E+01 0.00 S.42E+02 Ru-10S B.71E-03 0.00 3.44E-03 0.00 1.13E-01 0.00 S.33E+00 Ru-106 7.14E+01 0.00 9.03E+00 0.00 1.3BE+02 0.00 4.62E+03 Rh-10S 1.B4E+00 1.34E+00 B.BOE-01 0.00 S.6BE+00 0.00 2.13E+02 Ag-110m 1.20E+00 1.11E+00 6.61 E-01 0.00 2.19E+00 0.00 4.S4E+02 Sb-124 2.00E+03 3.77E+01 7.90E+02 4.B3E+00 0.00 1.SSE+03 S.66E+04 Sb-12S 1.61E+03 1.73E+01 3.22E+02 1.43E+00 0.00 1.6BE+OS 1.42E+04 Sb-126 7.B2E+02 1.S9E+01 2.B2E+02 4.79E+00 No Data 4.79E+02 6.39E+04 Te-12Sm 1.27E+04 4.60E+03 1.70E+03 3.B1 E+03 S.16E+04 0.00 S.06E+04 Te-127m 3.22E+04 1.1SE+04 3.93E+03 B.23E+03 1.31E+OS 0.00 1.0BE+OS Te-127 B.B9E+01 3.19E+01 1.92E+01 6.S9E+01 3.62E+02 0.00 7.01E+03 Te-129m S.40E+04 2.01E+04 B.S4E+03 1.BSE+04 2.2SE+OS 0.00 2.72E+OS Te-129 B.B9E-OS 3.34E-OS 2.17E-OS 6.B2E-OS 3.74E-04 0.00 6.71E-OSi Te-131m 4.76E+03 2.33E+03 1.94E+03 3.69E+03 2.36E+04 0.00 2.31E+Osl Te-131 4.32E-16 1.BOE-16 1.36E-16 3.SSE-16 1.B9E-1S 0.00 6.12E-17 Te-132 9.7SE+03 6.31E+03 S.92E+03 6.97E+03 6.0BE+04 0.00 2.9BE+05i 1-130 9.44E+00 2.7BE+01 1.10E+01 2.36E+03 4.34E+01 0.00 2.40E+01 !

1-131 1.B6E+02 2.66E+02 1.52E+02 B.71E+04 4.56E+02 0.00 7.01E+01 1-132 7.02E-03 1.BBE-02 6.57E-03 6.S7E-01 2.99E-02 0.00 3.53E-03 1

1-133 3.06E+01 S.33E+01 1.62E+01 7.B3E+03 9.30E+01 0.00 4.79E+01 i 1-134 2.91E-OB 7.92E-OB 2.B3E-OB 1.37E-06 1.26E-07 0.00 6.90E-11 1-135 1.71E+00 4.49E+00 1.66E+00 2.96E+02 7.20E+00 0.00 5.07E+00 Cs-134 2.99E+04 7.11E+04 5.B1E+04 0.00 2.30E+04 7.64E+03 1.24E+03 Cs-136 2.96E+03 1.17E+04 B.42E+03 0.00 6.S1E+03 B.92E+02 1.33E+03 Cs-137 3.B3E+04 5.24E+04 3.43E+04 0.00 1.7BE+04 S.92E+03 1.01E+03 Cs-13B 9.12E-13 1.BOE-12 B.92E-13 0.00 1.32E-12 1.31E-13 7.6BE-1Bi Ba-139 5.64E-OS 4.02E-OB 1.65E-06 0.00 3.76E-OB 2.28E-OB 1.00E-04 Ba-140 1.86E+03 2.34E+00 1.22E+02 0.00 7.95E-01 1.34E+00 3.B3E+03 Ba-141 0.00 0.00 0.00 0.00 0.00 0.00 0.00 8a-142 0.00 0.00 0.00 0.00 0.00 0.00 0.00 La-140 9.93E-02 S.01E-02 1.32E-02 0.00 0.00 0.00 3.6BE+03i La-142 2.19E-07 9.96E-OB 2.48E-OB 0.00 0.00 0.00 7.27E-04 Ce-141 4.40E+00 2.9BE+00 3.38E-01 0.00 1.3BE+00 0.00 1.14E+04 Ce-143 4.77E-01 3.53E+02 3.91E-02 0.00 1.55E-01 0.00 1.32E+04 Ce-144 2.34E+02 9.79E+01 1.26E+01 0.00 5.80E+01 0.00 7.91E+04 Pr-143 S.33E-01 2.14E-01 2.64E-02 0.00 1.23E-01 0.00 2.33E+03 Pr-144 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Nd-147 3.59E-01 4.1SE-01 2.48E-02 0.00 2.43E-01 0.00 1.99E+03

~

W-187 1.47E+02 1.23E+02 4.30E+01 0.00 0.00 4 Np-239 2.1SE-02 2.11E-03 1.17E-03 0.00 6.6 0.00 4 2-37 Version 24 01/10

FNP*ODCM 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2.5.1 Thirty-One Day Dose Projections In order to meet the requirements for operation of the LIQUID RADWASTE TREATMENT SYSTEM (see Section 2.1.4), dose projections must be made at least once each 31 days; this applies during periods in which a discharge to UNRESTRICTED AREAS of liquid effluents containing radioactive materials occurs or is expected.

Projected 31-day doses to individuals due to liquid effluents may be determined as follows:

(2.17) where:

= the projected dose to the total body or organ 't, for the next 31 days of liquid releases.

= the cumulative dose to the total body or organ 'to for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter).

= the anticipated dose contribution to the total body or any organ 't, due to any planned activities during the next 31-day period, if those activities will result in liquid releases that are in addition to routine liquid effluents. If only routine liquid effluents are anticipated, Dta may be set to zero.

2.5.2 Dose Projections for Specific Releases Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For individual dose projections due to liquid releases, follow the methodology of Section 2.4, using sample analysis results for the source to be released, and parameter values expected to exist during the release period.

2-38 Version 24 01/10

FNP*ODCM 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS The following symbolic terms are used in the presentation of liquid effluent calculations in the subsections above.

Section of Term Definition Initial Use Ap = the adjustment factor used in calculating the effluent monitor setpoint for liquid release pathway p: the ratio of the assured dilution to the required dilution [unitless]. 2.3.2.2 ADF = the assured dilution factor for a planned release [unitless]. 2.3.2.2 AFp = the dilution allocation factor for liquid release pathway p [unitless]. 2.3.2.2 Ait = the site-related adult ingestion dose commitment factor, for the total body or for any organ 't, due to identified radionuclide i [(mrem

mL) I (h

IlCi)]. The values of Aft are listed in Table 2-8. 2.4.1 Biv = the crop to soil concentration factor applicable to radionuclide i,

[(pCi/kg garden vegetation)/(pCi/kg soli)]. Values are listed in Table 2-6. 2.4.3 BFi = the bioaccumulation factor for radionuclide i for freshwater fish

[(pCi/kg) I (pCi/L)]. Values are listed in Table 2-6. 2.4.2 c = the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line, prior to dilution and subsequent release [IlCi/mL]. 2.3.2.1 cp = the calculated effluent radioactivity monitor setpoint for liquid release pathway p [IlCi/mL]. 2.3.2.2 Ca = the gross concentration of alpha emitters in the liquid waste as measured in the applicable composite sample [IlCi/mL]. 2.3.2.2 C ECL = the Effluent Concentration Limit stated in 10 CFR 20, Appendix B, Table 2, Column 2 [IlCi/mL]. 2.3.2.1 Cf = the concentration of Fe-55 in the liquid waste as measured in the applicable composite sample [IlCi/mL]. 2.3.2.2 Cg = the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed on the applicable pre-release waste sample [IlCi/mL]. 2.3.2.2 Cj = the measured concentration of radionuclide i in a sample of liquid effluent [IlCi/mL]. 2.3.2.2 Cil = the average concentration of radionuclide i in undiluted liquid effluent during time period I lIlCi/mL]. 2.4.1 2-39 Version 24 01/10

FNP-ODCM Section of Term Definition Initial Use C ir = the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution stream [IlCilmL]. 2.3.2.2 Cs = the concentration of strontium radioisotope s (Sr-89 or Sr-90) in the liquid waste as measured in the applicable composite sample

[IlCi/mL]. 2.3.2.2 Ct = the concentration of H-3 in the liquid waste as measured in the applicable composite sample [IlCi/mL]. 2.3.2.2 CFjv = the concentration factor for radionuclide i in irrigated garden vegetation [(pCi/kg) I (pCi/L)]. 2.4.2 Dw = the dilution factor from the near field of the discharge structure to the potable water intake location [unitless]. 2.4.2 Dt = the cumulative dose commitment to the total body or to any organ 't, due to radioactivity in liquid effluents released during a given time period [mrem]. 2.4.1 D'ta = the anticipated dose contribution to the total body or any organ t, due to any planned activities during the next 31-day period [mrem]. 2.5.1 D,c = the cumulative dose to the total body or organ 't, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrem]. 2.5.1 Dw = the projected dose to the total body or organ 't, for the next 31 days of liquid releases [mrem]. 2.5.1 DFit = the dose conversion factor for radionuclide i for adults, in organ 't

[mrem/pCi]. Values are listed in Table 2-7. 2.4.2 ECLj = the liquid Effluent Concentration Limit for radionuclide i from 10 CFR Part 20, Appendix B, Table 2, Column 2 [IlCi/mL]. 2.3.2.2 f = the effluent flowrate at the location of the radioactivity monitor [gpm]. 2.3.2.1 fap = the anticipated actual discharge flowrate for a planned release from liquid release pathway p [gpm]. 2.3.2.2 fl = the fraction of the year that garden vegetation is irrigated [unitless]. 2.4.3 fmp = the maximum permissible effluent discharge flowrate for release pathway p [gpm]. 2.3.2.2 fr = the effluent discharge flowrate of release pathway r [gpm]. 2.3.2.2 2-40 Version 24 01/10

FNP*ODCM Section of Term Definition Initial Use ft = the average undiluted liquid waste f10wrate actually observed during the period of a liquid release [gpm]. 2.4.1 F = the dilution stream f10wrate which can be assured prior to the release point to the UNRESTRICTED AREA [gpm]. 2.3.2.1 Fd = the entire assured dilution f10wrate for the plant site during the release period [gpm]. 2.3.2.2 Fdp = the dilution f10wrate allocated to release pathway p [gpm]. 2.3.2.2 FI = the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA [unitless]. 2.4.1 Ft = the average dilution stream f10wrate actually observed during the period of a liquid release [gpm}. 2.4.1

= the average irrigation rate during the growing season [L/(m 2*h}]. 2.4.3 Lv = the water content of leafy garden vegetation edible parts [Ukg]. 2.4.3 M = the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage

[unitless]. 2.4.3 p = the effective surface density of soil [kglm 2]. 2.4.3 r = the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation. 2.4.3 RDF = the required dilution factor: the minimum ratio by which liquid effluent must be diluted before reaching the UNRESTRICTED AREA, in order to ensure that the limits of Section 2.1.2 are not exceeded [unitless]. 2.3.2.2 RDFy = the RDF for a liquid release due only to its concentration of gamma-emitting radionuclides [unitless]. 2.3.2.2 RDFny = the RDF for a liquid release due only to its concentration of non-gamma-emitting radionuclides [unitless]. 2.3.2.2 SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement [unitless]. 2.3.2.2 t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration. 2.5.1 tb = the period of long-term buildup of activity in soil [h]. 2.4.3 te = the period of leafy garden vegetation exposure during the growing season [h}. 2.4.3 2-41 Version 24 01/10

FNP-ODCM Section of Term Definition Initial Use tf = the transit time from release to receptor for fish consumption [h]. 2.4.2 th = the time between harvest of garden vegetation and human consumption [h). 2.4.3 tw = the transit time from release to receptor for potable water consumption [h). 2.4.2 TF = the tolerance factor selected to allow flexibility in the establishment of a practical monitor setpoint which could accommodate effluent releases at concentrations higher than the ECl values stated in 10 CFR 20, Appendix B, Table 2, Column 2 [unitless]; the tolerance factor must not exceed a value of 10. 2.3.2.2 Uf = the adult rate of fish consumption [kg/y]. 2.4.2 Uy = the adult consumption rate for irrigated garden vegetation [kg/y). 2.4.2 Uw = the adult drinking water consumption rate applicable to the plant site

[L/y]. 2.4.2 Yy = the areal density (agricultural productivity) of leafy garden vegetation

[kg/m 2 ). 2.4.3 Z = the applicable dilution factor for the receiving water body, in the near field of the discharge structure, during the period of radioactivity release [unitless). 2.4.1

~tl = the length of time period I, over which Cil and FI are averaged for liquid releases [h). 2.4.1 A.ei = the effective removal rate for activity deposited on crop leaves

[h-1]. 2.4.3

~ = the decay constant for radionuclide i [h-1). 2.4.2

'Aw = the rate constant for removal of activity from plant leaves by weathering [h- 1]. 2.4.3 2-42 Version 24 01/10

FNP-ODCM CHAPTER 3 GASEOUS EFFLUENTS 3.1 LIMITS OF OPERATION The following Limits of Operation implement requirements established by Technical Specifications Section 5.0. Terms printed in all capital letters are defined in Chapter 10.

3.1.1 Gaseous Effluent Monitoring Instrumentation Control In accordance with Technical Specification 5.5.4.a, the radioactive gaseous el'nuent monitoring instrumentation channels shown in Table 3-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Section 3.1.2.a are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with Section 3.3.

3.1.1.1 Applicability These limits apply as shown in Table 3-1.

3.1.1.2 Actions With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, declare the channel inoperable, or restore the setpoint to a value that will ensure that the limits of Section 3.1.2.a are met.

With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3-1. Restore the inoperable instrumentation to OPERABLE status within 30 days and, if unsuccessful, explain in the next Radioactive Effluent Release Report pursuant to Section 7.2 why this inoperability was not corrected in a timely manner.

This control does not affect shutdown requirements or MODE changes.

3.1.1.3 Surveillance Requirements Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION, and CHANNEL OPERATIONAL TEST (COT) operations at the frequencies shown in Table 3-2.

3-1 Version 24 01/10

FNP-ODCM 3.1.1.4 Basis 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 AlarmfTrip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section 3.3 to ensure that the alarm/trip will occur prior to exceeding the limits of Section 3.1.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

3-2 Version 24 01/10

FNP-ODCM Table 3*1 Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirements!)

Minimum Channels Instrument OPERABLE Applicability ACTION

1. Steam Jet Air Ejector Noble Gas Activity Monitor (RE-15) 1 MODES 1,2,3,4 37

2. Plant Vent Stack

a. Noble Gas Activity Monitor (RE-14 or RE-22) 1 At all times 37a

b. Iodine Sampler 1 At all times 39 c Particulate Sampler 1 At all times 39

d. Flowrate Monitor 1 At all times 36

3. GASEOUS RADWASTE TREATMENT SYSTEM Noble Gas Activity Monitor (RE-14), with Alarm and Automatic Termination of Release 1 Atal! times 35

a. For continuous releases.

b. All requirements in this table apply to each unit.

3-3 Version 24 01/10

FNP*ODCM Table 3*1 (contd) Notation for Table 3 ACTION Statements ACTION 35 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment provided that prior to initiating the release:

a. At least two independent samples of the tank's contents are analyzed, and

b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and (1) Verify the manual portion of the computer input for the release rate calculations performed on the computer, or (2) Verify the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 36 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 37 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab samples are taken at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and these samples are analyzed for gross activity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 39 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are continuously collected with auxiliary sampling equipment as required in Table 3-3.

3-4 Version 24 01/10

FNP-ODCM Table 3-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements d

Surveillance Requirements CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST MODES c

1. Steam Jet Air Ejector Noble Gas Activity Monitor (RE-15) D M Rb 0"(2) 1,2,3,4

2. Plant Vent Stack

a. Noble Gas Activity Monitor D M Rb 0"(1.2)

All RE-14 D M Rb Oa(2)

All RE-22

b. Iodine Sampler W NA NA NA All

c. Particulate Sampler W NA NA NA All

d. Flowrate Monitor D NA R 0 All

a. In addition to the basic functions of a CHANNEL OPERATIONAL TEST (Section 10.2):

(1) The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room annunciation occur if any of the following conditions exists:

(a) Instrument indicates measured levels above the alarm/trip setpoint; (b) Loss of control power; or (c) Loss of instrument power.

(2) The CHANNEL OPERATIONAL TEST shall also demonstrate that control room annunciation occurs if any of the following conditions exists:

(a) Instrument indicates a downscale failure; or (b) Instrument controls not set in the OPERATE mode.

b. The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology, or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. For any subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

c. MODES in which surveillance is required. "All" means "At all times."

All requirements in this table apply to each unit.

3-5 Version 24 01/10

FNP-ODCM 3.1.2 Gaseous Effluent Dose Rate Control In accordance with Technical Specifications 5.5.4.c and 5.5.4.g, the licensee shall conduct operations so that the dose rates due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 10-1) are limited as follows:

a. For noble gases: Less than or equal to a dose rate of 500 mrem/y to the total body and less than or equal to a dose rate of 3000 mrem/y to the skin, and

b. For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/y to any organ.

3.1.2.1 Applicability This limit applies at all times.

3.1.2.2 Actions With a dose rate due to radioactive material released in gaseous effluents exceeding the limit stated in Section 3.1.2, immediately decrease the release rate to within the stated limit.

This control does not affect shutdown requirements or MODE changes.

3.1.2.3 Surveillance Requirements The dose rates due to radioactive materials in areas at or beyond the SITE BOUNDARY due to releases of gaseous effluents shall be determined to be within the above limits, in accordance with the methods and procedures in Section 3.4.1, by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 3-3.

3.1.2.4 Basis This control is provided to ensure that gaseous effluent dose rates will be maintained within the limits that historically have provided reasonable assurance that radioactive material discharged in gaseous effluents will not result in a dose to a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, exceeding the limits specified in Appendix I of 10 CFR Part 50, while allowing operational flexibility for effluent releases. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be suffiCiently low to compensate for any increase in the atmospheriC diffusion factor above that for the SITE BOUNDARY.

The dose rate limit for lodine-131, lodine-133, tritium, and radionuclides in particulate form with half lives greater than 8 days specifically applies to dose rates to a child via the inhalation pathway.

This control applies to the release of gaseous effluents from all reactors at the site.

3-6 Version 24 01/10

FNP*ODCM Table 3-3 Radioactive Gaseous Waste Sampling and Analysis Program Sampling and Analysis Requirementsa,b MINIMUM DETECTABLE Gaseous Minimum CONCENTRATION Release Sampling Analysis Type of Activity (MDC)

Type FREQUENCY FREQUENCY Analysis (I-LCi/mL)

P P PRINCIPAL Waste Gas Each Tank Each Tank GAMMA Decay Tank Grab Sample EMITTERS 1 E-4 Noble Gas pc PRINCIPAL pc Containment Each Purge GAMMA Each Purge EMITTERS 1 E-4 Purge Grab Sample (batch release)

M H-3 1 E-6 Condenser Noble Gas Steam Jet PRINCIPAL Air GAMMA Mc,d,1 Me Ejector, EMITTERS 1 E-4 Plant Vent Grab Sample Stack H-3 1 E-6 We CONTINUOUS9 Charcoal or Charcoal or 1-131 1 E-12 Silver Silver Zeolite Plant Vent Zeolite Sample 1-133 1 E-10 Stack, We Containment PRINCIPAL Purge Particulate GAMMA (continuous CONTINUOUS9 Sample EMITTERS 1 E-11 purge) M COMPOSITE Particulate CONTINUOUS9 Sample Gross Alpha 1 E-11 Q

COMPOSITE Particulate CONTINUOUS9 Sample Sr-89, Sr-90 1 E-11 Noble Gases Noble Gas (Gross Beta and CONTINUOUS9 Monitor Gamma) 1 E-6 3-7 Version 24 01/10

FNP*ODCM Table 3-3 (cont'd) Notation for Table 3-3

a. All requirements in this table apply to each unit. Deviation from the MDC requirements of this table shall be reported in accordance with Section 7.2. Deviation from the composite sampling requirements of this table shall be reported in accordance with Section 7.2.

b. Terms printed in all capital letters are defined in Chapter 10.

c. Sampling and analyses shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one-hour period. This requirement does not apply if (1) analysis shows that measured DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the plant vent, containment purge, and steam jet air ejector noble gas monitors show that effluent activity has not increased more than a factor of 3.

d. Tritium grab samples shall be taken from the plant vent stack at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling canal is flooded.

e. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a one-hour period, and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding MDC may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that measured DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the plant vent, containment purge, and steam jet air ejector noble gas monitors show that effluent activity has not increased more than a factor of 3.

f. Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is in the spent fuel pool.

g. The ratio of the sample flow rate to the sampled stream f10wrate shall be known for the time period covered by each dose or dose rate calculation made in accordance with controls specified in Sections 3.1.2, 3.1.3, and 3.1.4.

3-8 Version 24 01/10

FNP*ODCM 3.1.3 Gaseous Effluent Air Dose Control In accordance with Technical Specifications 5.5.4.e and 5.5.4.h, the air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

3.1.3.1 Applicability This limit applies at all times.

3.1.3.2 Actions With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to 10 CFR 50.4, a Special Report which identifies the cause(s) for exceeding the limit(s);

defines the corrective actions that have been taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases of radioactive noble gases in gaseous effluents will be in compliance with the limits of Section 3.1.3.

This control does not affect shutdown requirements or MODE changes.

3.1.3.3 Surveillance Requirements Cumulative air dose contributions from noble gas radionuclides released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.2 at least once per 31 days.

3.1.3.4 Basis This control is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. Section 3.1.3 implements the guides set forth in Section II.B of Appendix I. The ACTION statements in Section 3.1.3.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I, assuring that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance requirements in Section 3.1.3.3 implement the requirements in Section III.A of Appendix I, which require that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section 3.4.2 for calculating the doses due to the actual releases of noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). The equations in Section 3.4.2 provided for determining the air doses at the SITE BOUNDARY are based upon the historical annual average atmospheric conditions.

3-9 Version 24 01/10

FNP*ODCM 3.1.4 Control on Gaseous Effluent Dose to a MEMBER OF THE PUBLIC In accordance with Technical Specifications 5.5.4.e and 5.5.4.i, the dose to a MEMBER OF THE PUBLIC from 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 0-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ, and

b. During any calendar year: Less than or equal to 15 mrem to any organ.

3.1.4.1 Applicability This limit applies at all times.

3.1.4.2 Actions With the calculated dose from the release of 1-131,1-133, tritium, or radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to 10 CFR 50.4, a Special Report which identifies the cause(s) for exceeding the limit; defines the corrective actions that have been taken to reduce the releases of radioiodines and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents; and defines proposed corrective actions to assure that subsequent releases will be in compliance with the limits stated in Section 3.1.4.

This control does not affect shutdown requirements or MODE changes.

3.1.4.3 Surveillance Requirements Cumulative organ dose contributions to a MEMBER OF THE PUBLIC from 1-131,1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.3 at least once per 31 days.

3.1.4.4 Basis This control is provided to implement the requirements of Section II.C, III.A and IV.A of Appendix I, 10 CFR Part 50. The limits stated in Section 3.1.4 are the guides set forth in Section II.C of Appendix I. The ACTION statements in Section 3.1.4.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The calculational methods specified in the Surveillance Requirements of Section 3.1.4.3 implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The calculational methods in Section 3.4.3 for calculating the doses due to the actual releases of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Version 24 01/10

FNP-ODCM Regulatory Guide 1.111 (Reference 5). These equations provide for determining the actual doses based upon the historical annual average atmospheric conditions. The release specifications for radioiodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy garden vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.

3.1.5 GASEOUS RADWASTE TREATMENT SYSTEM Control In accordance with Technical Specification 5.5.4.f, the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous wastes prior to their discharge when the projected air doses due to gaseous effluent releases, from each reactor unit, to areas at and beyond the SITE BOUNDARY (see Figure 10-1) would exceed 0.2 mrad for gamma radiation or 0.4 mrad for beta radiation in 31 days. The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous wastes prior to their discharge when the projected doses due to gaseous effluent releases, from each reactor unit, to areas beyond the SITE BOUNDARY (see Figure 10-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC in 31 days.

3.1.5.1 Applicability These limits apply at all times.

3.1.5.2 Actions With gaseous waste being discharged without treatment and in excess of the limits in Section 3.1.5, prepare and submit to the Nuclear Regulatory Commission within 30 days, pursuant to 10 CFR 50.4, a Special Report which includes the following information:

a. Identification of the inoperable equipment or subsystem and the reason for inoperability,

b. Action(s) taken to restore the inoperable equipment to OPERABLE status, and

c. Summary description of action(s) taken to prevent a recurrence.

This control does not affect shutdown reqUirements or MODE changes.

3.1.5.3 Surveillance Requirements Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days, in accordance with Section 3.5.1, when the GASEOUS RADWASTE TREATMENT SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.

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FNP-ODCM The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE:

a. by meeting the controls of Sections 3.1.2, and either 3.1.3 (for the GASEOUS RADWASTE TREATMENT SYSTEM) or 3.1.4 (for the VENTILATION EXHAUST TREATMENT SYSTEM), or

b. by operating the GASEOUS RADWASTE TREATMENT SYSTEM eqUipment and the VENTILATION EXHAUST TREATMENT SYSTEM equipment for at least 15 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

3.1.5.4 Basis The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable." This control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section 11.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of these systems were specified as a suitable fraction of the dose design opjectives set forth in Section 11.8 and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents.

3.1.6 MAJOR CHANGES TO THE GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEM AND THE VENTILATION EXHAUST TREATMENT SYSTEM Licensee-initiated MAJOR CHANGES to the GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM:

a. Shall be reported to the Nuclear Regulatory Commission in the Radioactive Effluent Release Report for the period in which the change was implemented, in accordance with Section 7.2.2.7.

b. Shall become effective upon review by the Plant Review Board and approval by the Vice President-Plant.

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3.2 GASEOUS RADWASTE TREATMENT SYSTEM At the Farley Nuclear Plant, there are six designated points where radioactivity may be released to the atmosphere in gaseous discharges: the Unit 1 and Unit 2 Plant Vent Stacks; the Unit 1 and Unit 2 Turbine Building Vents (steam jet air ejectors); and the Unit 1 and Unit 2 Integrated Leak Rate Test (lLRT) Vents. Of these six, only four are routine release pathways, since ILRT Vent releases are performed only infrequently.

Figure 3-1 gives schematic diagrams of the Waste Gas Treatment Systems and the Ventilation Systems (Reference 7). Discharges from the two reactor units are separated, with no shared systems. In each unit, Containment Purge and Waste Gas Decay Tank effluents are discharged through the respective Plant Vent, and are treated as contributions to the on-going Plant Vent CONTINUOUS release. Although Waste Gas Decay Tank effluents are released via the Plant Vent Stack, they are tracked separately and accounted for as BATCH releases.

Table 3-4 summarizes the release height and release type characteristics of the various release pathways and source streams. Chapter 8 discusses the calculation of atmospheriC dispersion parameters using the ground-level and mixed-mode (Le., split-wake) models.

As established in Section 3.1.1, gaseous effluent monitor setpoints are required for the noble gas monitors on the two Plant Vents and the two Turbine Building Vents (steam jet air ejectors).

Waste Gas Treatment System discharges are not monitored separately during release, but are sampled prior to release and are monitored by the downstream Plant Vent monitors during release. ILRT discharges are not monitored during release, but are sampled prior to release; the ILRT Vent may be assigned an appropriate allocation factor during the release period, and dose calculations may be based on estimates of the activity concentration and the volume of air released. Sampling and analysis of both these release pathways must be sufficient to ensure that the gaseous effluent dose limits specified in Section 3.1.3 and Section 3.1.4 are not exceeded.

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FNP*ODCM Waste Gas Decay Tanks Waste Gas Compressor PLANT VENT STACK I

. __._.................__.......f \ ........

MONITORS CONTAINMENT AUX.

CONTAIN* PURGE BLDG.

MENT EXHAUST PLENUM MONITOR MONITOR TURBINE TURBINE BUILDING BUILDING VENT (STEAM JET AIR EJECTOR)

Figure 3*1 Schematic Diagram of the Routine Release Sources and Release Points (Typical of Both Units) 3-14 Version 24 01/10

FNP-ODCM 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3.3.1 General Provisions Regarding Noble Gas Monitor Setpoints Noble gas radioactivity monitor setpoints calculated in accordance with the methodology presented in this section are intended to ensure that the limits of Section 3.1.2.a are not exceeded. They will be regarded as upper bounds for the actual high alarm setpoints. That is, a lower high alarm setpoint may be established or retained on the monitor, if desired.

Intermediate level setpoints should be established at an appropriate level to give sufficient warning prior to reaching the high alarm setpoint.

If no release is planned for a given pathway, or if there is no detectable activity in the gaseous stream being evaluated for release, the setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should a significant inadvertent release occur.

Section 3.1.1 establishes the requirements for gaseous effiuent monitoring instrumentation, and Section 3.2 describes the VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS RADWASTE TREATMENT SYSTEM. From those sections, it can be seen that certain monitors are located on final release pathways, that is, streams that are being monitored immediately before being discharged from the plant; the setpoint methodology for these monitors is presented in Section 3.3.2. Other monitors are located on source streams, that is, streams that merge with other streams prior to passing a final monitor and being discharged; the setpoint methodology for these monitors is presented in Section 3.3.3. Table 3-4 identifies which of these setpoint methodologies applies to each monitor. Some additional monitors with special setpoint requirements are discussed in Section 3.3.5.

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FNP-ODCM Table 3-4 Applicability of Gaseous Monitor Setpoint Methodologies Final Release Pathways with no Monitored Source Streams Release Elevation: Ground-level Unit 1 or Unit 2 Turbine Building Vent Release Type: CONTINUOUS Monitor: 1RE-15/2RE-15 Setpoint Method: Section 3.3.2 Maximum Flowrate: 1060 cfm (5.00 E+05 mUs)

Unit 1 or Unit 2 ILRT Vent Release Type: BATCH Monitor: None Setpoint Method: None Maximum Flowrate: Release-dependent Final Release Pathways with One or More Monitored Source Streams Release Elevation: Mixed-Mode Unit 1 or Unit 2 Plant Vent Stack Release Type: CONTINUOUS Monitor: 1RE-14/2RE-14, and 1RE-22 I 2RE-22 Setpoint Method: Section 3.3.2 Maximum Flowrate: 150,000 cfm (7.08 E+07 mUs)

Source Stream: Unit 1 or Unit 2 Containment Purge Release Type: CONTINUOUS or BATCH Monitor: 1RE-24 12RE-24 Setpoint Method: Section 3.3.3 is optional. See Section 3.3.5.

Maximum Flowrate: Release-dependent Source Stream: Unit 1 or Unit 2 Waste Gas Decay Tanks Release Type: BATCH Monitor: None Setpoint Method: None Maximum Flowrate: Release-dependent (X/Q)vb Values for Use in Setpoint Calculations Ground-Level Releases: 4.87 x 10-5 s/m 3 [S Sector]

Mixed-Mode Releases: 1.08 x 10-6 s/m 3 [SSE Sector]

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FNP-ODCM 3.3.2 Setpoint for the Final Noble Gas Monitor on Each Release Pathway 3.3.2.1 Overview of Method Gaseous effluent radioactivity monitors are intended to alarm prior to exceeding the limits of Section 3.1.2.a. Therefore, their alarm setpoints are established to ensure compliance with the following equation:

AG.SF'X'~

c = the lesser of

{AG*SF*X*R k (3.1) where:

c = the setpoint, in IlCi/mL, of the radioactivity monitor measu ring the concentration of radioactivity in the effluent line prior to release. The setpoint represents a concentration which, if exceeded, could result in dose rates exceeding the limits of Section 3.1.2.a at or beyond the SITE BOUNDARY.

AG = an administrative allocation factor applied to divide the release limit among all the gaseous release pathways at the site.

SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement.

x = the noble gas concentration for the release under consideration.

Rt = the ratio of the dose rate limit for the total body, 500 mrem/y, to the dose rate to the total body for the conditions of the release under consideration.

the ratio of the dose rate limit for the skin, 3000 mrem/y, to the dose rate to the skin for the conditions of the release under consideration.

Equation (3.1) shows the relationships of the critical parameters that determine the setpoint.

However, in order to apply the methodology presented in the equation to a mixture of noble gas radionuclides, radionuclide-specific concentrations and dose factors must be taken into account under conditions of maximum flowrate for the release point and annual average meteorology.

The basic setpoint method presented below is applicable to the radioactivity monitor nearest the point of release for the release pathway. For monitors measuring the radioactivity in source streams that merge with other streams prior to subsequent monitoring and release, the modifications presented in Section 3.3.3 must be applied.

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FNP-ODCM 3.3.2.2 Setpoint Calculation Steps Step 1: Determine the concentration, Xiv, of each noble gas radionuclide i in the gaseous stream v being considered for release, in accordance with the sampling and analysis requirements of Section 3.1.2. Then sum these concentrations to determine the total noble gas concentration, L: Xiv' Step 2: Determine Rio the ratio of the dose rate limit for the total body, 500 mrem/y, to the total body dose rate due to noble gases detected in the release under consideration, as follows:

R = 500 (3.2)

I ( XlQ)vb L: [Iei .

i QJ where:

500 = the dose rate limit for the total body, 500 mrem/y.

(XlQ)vb = the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v.

Table 3-4 includes an indication of what release elevation is applicable to each release pathway; release elevation determines the appropriate value of (XlQ)vb.

= the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mrem/y)/(!lCi/m3 ), from Table 3-5.

= the release rate of noble gas radionuclide i from the release pathway under consideration, in !lCi/s, calculated as the product of ~v and fav, where:

~v = the concentration of noble gas radionuclide i for the particular release, in !lCi/mL.

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FNP-ODCM fay = the maximum anticipated f10wrate for release pathway v during the period of the release under consideration, in mUs.

Step 3: Determine Rk , the ratio of the dose rate limit for the skin, 3000 mrem/y, to the skin dose rate due to noble gases detected in the release under consideration, as follows:

3000 (3.3) where:

3000 = the dose rate limit for the skin, 3000 mrem/y.

= the skin dose factor due to beta emissions from noble gas radionuclide i, in (mrem/y)/(~Ci/m\ from Table 3-5.

= the air dose factor due to gamma emissions from noble gas radionuclide i, in (mrad/y)/(~Ci/m\ from Table 3-5.

1.1 = the factor to convert air dose in mrad to skin dose in mrem.

All other terms were defined previously.

Step 4: Determine the maximum noble gas radioactivity monitor setpoint concentration.

Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the limits of Section 3.1.2.a will not be exceeded. Because the radioactivity monitor responds primarily to radiation from noble gas radionuclides, the monitor setpoint Cnv (in ~Ci/mL) is based on the concentration of all noble gases in the waste stream, as follows:

AGv *SF* .t....J

"" X IV .1)

.1'1 env = the lesser of i (3.4)

{

AGv ' SF* LXiv *Rk 3-19 Version 24 01/10

FNP*ODCM where:

cnv = calculated setpoint, in IlCi/mL, for the noble gas monitor serving gaseous release pathway v.

AG v = the administrative allocation factor for gaseous release pathway v, applied to divide the release limit among all the gaseous release pathways at the site.

The allocation factor may be assigned any value between 0 and 1, under the condition that the sum of the allocation factors for all simultaneously-active final release pathways at the entire plant site does not exceed 1. Alternative methods for determination of AG v are presented in Section 3.3.4.

SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement. The value for the safety factor must be between 0 and 1. A value of 0.5 is reasonable for gaseous releases; a more precise value may be developed if desired.

= the measured concentration of noble gas radionuclide i in gaseous stream v, as defined in Step 1 , in IlCilmL.

The values of Rt and Rk to be used in the calculation are those which were determined in Steps 2 and 3 above.

Step 5: Determine whether the release is permissible, as follows:

If C nv ~ 2: Xiv ' the release is permissible. However, if C nv is within about 10 percent of 2: Xiv, i

it may be impractical to use this value of cnv . This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to adjust the allocation of the limits among the active release points. The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions.

If C nv < 2: x iv , the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.

3.3.2.3 Use of the Calculated Setpoint The setpoint calculated above is in the units IlCi/mL. The monitor actually measures a count rate that includes background, so that the calculated setpoint must be converted accordingly:

(3.5) where:

= the monitor setpoint as a count rate.

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FNP-ODCM Ev = the monitor calibration factor, in count rate/(IlCi/mL). Monitor calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to effluent stream concentrations measured by sample analysis.

= the monitor background count rate. In all cases, monitor background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value. Contributions to the monitor background may include any or all of the following factors:

ambient background radiation, plant-related radiation levels at the monitor location (which may change between shutdown and power conditions), and internal background due to contamination of the monitor's sample chamber.

The count rate units for c*nv , E" and Bv in equation (3.5) must be the same, cpm or cps.

3.3.3 Setpoints for Noble Gas Monitors on Effluent Source Streams Table 3-4 lists certain gaseous release pathways as being source streams. As may be seen in the figures of Section 3.2, these are streams that merge with other streams, prior to passing a final radioactivity monitor and being released. Unlike the final monitors, the source stream monitors measure radioactivity in effluent streams for which flow can be terminated; therefore, the source stream monitors have control logic to terminate the source stream release at the alarm setpoint.

3.3.3.1 Setpoint of the Monitor on the Source Stream Step 1: Determine the concentration Xis of each noble gas radionuclide i in source stream s (in /lCi/mL) according to the results of its required sample analyses

[see Section 3.1.2].

Step 2: Determine rt, the ratio of the dose rate limit for the total body, 500 mrem/y, to the total body dose rate due to noble gases detected in the source stream under consideration. Use the Xis values and the maximum anticipated source stream flow rate fas in equation (3.2) to determine the total body dose rate for the source stream, substituting rt for Rt.

The SITE BOUNDARY relative dispersion value used in Steps 2 and 3 for the source stream is the same as the (X/Q>vt, that applies to the respective merged stream. This is because the (X/Q) value is determined by the meteorology of the plant site and the physical attributes of the release point, and is unaffected by whether or not a given source stream is operating.

Step 3: Determine rk, the ratio of the dose rate limit for the skin, 3000 mrem/y, to the skin dose rate due to noble gases detected in the source stream under consideration.

Use the Xis values and the maximum anticipated source stream flow rate fas in equation (3.3) (if the release is elevated) to determine the skin dose rate for the source stream, substituting rk for Rk.

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FNP*ODCM Step 4: Determine the maximum noble gas radioactivity monitor setpoint concentration, as follows:

AG' . SF . L.x;, . 1i ens = the lesser of i (3.6)

{ AG, . SF* L.XI's . rk i

where:

cns = the calculated setpoint (in /J.Ci/mL) for the noble gas monitor serving gaseous source stream s.

= the administrative allocation factor applied to gaseous source stream

s. For a given final release point v, the sum of all the AG s values for source streams contributing to the final release point must not exceed the release point's allocation factor AG v *

= the measured concentration of noble gas radionuclide i in gaseous source stream s, as defined in Step 1, in /J.Ci/mL.

The values of rt and rk to be used in the calculation are those which were determined in Steps 2 and 3 above. The safety factor, SF, was defined previously.

Step 5: Determine whether the release is permissible, as follows:

If cns ~ L:Xis the release is permissible. However, if cns is within about 10 percent of L:Xis it i ' i '

may be impractical to use this value of cns. This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to adjust the allocation of the limits among the active release points. The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions.

If Cns < L:Xis the release may not be made as planned. Consider the alternatives discussed i '

in the paragraph above, and calculate a new setpoint based on the results of the actions taken.

3.3.3.2 Effect on the Setpoint of the Monitor on the Merged Stream Before beginning a release from a monitored source stream, a setpoint must be determined for the source stream monitor as presented in Section 3.3.3.1. In addition, whether or not the source stream has its own effluent monitor, the previously-determined maximum allowable setpoint for the downstream final monitor on the merged stream must be redetermined. This is accomplished by repeating the steps of Section 3.3.2, with the following modifications.

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FNP-ODCM Modification 1: The new maximum anticipated flowrate of the merged stream is the sum of the old merged stream maximum flowrate, ( (fav)old), and the maximum flowrate of the source stream being considered for release (fas).

(3.7)

Modification 2: The new concentration of noble gas radionuclide i in the merged stream includes both the contribution of the merged stream without the source stream, and the source stream being considered for release.

(X) lVnew

= Va" )old * (Xiv)old

(.t')

+ las

Xis (3.8)

Vav nelV 3.3.4 Determination of Allocation Factors, AG When simultaneous gaseous releases are conducted, an administrative allocation factor must be applied to divide the release limit among the active gaseous release pathways. This is to assure that the dose rate limit for areas at and beyond the SITE BOUNDARY (see Section 3.1.2) will not be exceeded by simultaneous releases. The allocation factor for any pathway may be assigned any value between 0 and 1, under the following two conditions:

1. The sum of the allocation factors for all simultaneously-active final release paths at the plant site may not exceed 1.

2. The sum of the allocation factors for all simultaneously-active source streams merging into a given final release pathway may not exceed the allocation factor of that final release pathway.

Any of the following three methods may be used to assign the allocation factors to the active gaseous release pathways:

1. For ease of implementation, AG v may be equal for all release pathways:

1 AG = (3.9) v N where:

N= the number of simultaneously active gaseous release pathways.

2. AG v for a given release pathway may be selected based on an estimate ofthe portion of the total SITE BOUNDARY dose rate (from all simultaneous releases) that is contributed by the release pathway. During periods when a given building or release pathway is not subject to gaseous radioactive releases, it may be assigned an allocation factor of zero.

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FNP-ODCM

3. AG v for a given release pathway may be selected based on a calculation of the portion of the total SITE BOUNDARY dose rate that is contributed by the release pathway, as follows, CVQL L (KiQiv)

=

t AGv _-r~_-,-i_ _ _.....

(3.10)

[CVQ)rb ~ (KiQr) where:

(XlQ)vb = the annual average SITE BOUNDARY relative concentration applicable to the gaseous release pathway v for which the allocation factor is being determined, in s/m 3

Ki = the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mrem/y)/(IlCi/m\ from Table 3-5.

Qjv = the release rate of noble gas radionuclide i from release pathway v, in IlCi/s, calculated as the product of Xiv and fav, where:

Xiv = the concentration of noble gas radionuclide i applicable to the gaseous release pathway v for which the allocation factor is being determined, in IlCi/mL.

fav = the discharge flowrate applicable to gaseous release pathway v for which the allocation factor is being determined, in mUs.

(XlQ)rb = the annual average SITE BOUNDARY relative concentration applicable to 3

active gaseous release pathway r, in s/m

Qir = the release rate of noble gas radionuclide i applicable to active release pathway r, in IlCi/s, calculated as the product of X ir and far. where:

Xjr = the concentration of noble gas radionuclide i applicable to active gaseous release pathway r, in IlCi/mL.

far = the discharge flowrate applicable to active gaseous release pathway r, in mUs.

N = the number of simultaneously active gaseous release pathways (including pathway v that is of interest).

Note: Although equations (3.9) and (3.10) are written to illustrate the assignment of the allocation factors for final release pathways, they may also be used to assign allocation factors to the source streams that merge into a given final release pathway.

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FNP-ODCM 3.3.5 Setpoints for Noble Gas Monitors with Special Requirements The Farley Nuclear Plant operating philosophy treats the Waste Gas Decay Tank supply monitors (1/2 RE-013) and the Containment Purge monitors (1/2 RE-024) as process monitors, not effluent monitors. However, as a matter of information, the following may be noted regarding their setpoints:

For 1/2 RE-013, the alarm setpoint should be based on a concentration equivalent to no more than the Technical Requirements Manual limit for the maximum curie content of a Waste Gas Decay Tank. In converting the curie limit to an equivalent concentration at the location of RE-013, the maximum allowable Waste Gas Decay Tank pressure should be used.

For 1/2 RE--024, the alarm setpoint concentration may be arrived at in either of two ways. In the first method, the maximum setpoint concentration established by the Technical Specifications may be used. Alternatively, to provide early detection and termination of an abnormally high containment purge release, the [lower] setpoint concentration calculated according to Section 3.3.3 may be used.

3.3.6 Setpoints for Particulate and Iodine Monitors In accordance with Section 5.1.1 of NRC NUREG-0133 (Reference 1), the effluent controls of Section 3.1.1 do not require that the ODCM establish setpoint calculation methods for particulate and iodine monitors.

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FNP*ODCM 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3.4.1 Dose Rates at and Beyond the Site Boundary Because the dose rate limits for areas at and beyond the SITE specified in Section 3.1.2 are site limits applicable at any instant in time, the summations extend over all simultaneously active gaseous final release pathways at the plant site. Table 3-4 identifies the gaseous final release pathways at the plant site, and indicates the (X/Q)vb value for each.

3.4.1.1 Dose Rates Due to Noble Gases For the purpose of implementing the controls of Section 3.1.2.a, the dose rates due to noble gas radionuclides in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:

For total body dose rates:

(3.11 )

For skin dose rates:

(3.12) where:

DRt = the total body dose rate at the time of the release, in mrem/y.

DRk = the skin dose rate at the time of the release, in mrem/y.

qv = the release rate of noble gas radionuclide i, in IlCi/s, equal to the product of ftv and Xiv, where:

ftv = the actual average flowrate for release pathway v during the period of the release, in mLls.

All other terms were defined previously.

3-26 Version 24 01/10

FNP-ODCM 3.4.1.2 Dose Rates Due to lodine-131, lodine-133, Tritium, and Radionuclides in Particulate Form with Half-Lives Greater than 8 Days.

For the purpose of implementing the controls of Section 3.1.2.b, the dose rates due to lodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:

(3.13) where:

DRo = the dose rate to organ 0 at the time of the release, in mrem/y.

= the site-specific dose factor for radionuclide i and organ 0, in (mrem/y)/(IlCi/m3). Since the dose rate limits specified in Section 3.1.2.b apply only to the child age group exposed to the inhalation pathway, the values of Pio may be obtained from Table 3-9, "Raipj for Inhalation Pathway, Child Age Group."

= the release rate of radionuclide i from gaseous release pathway v, in IlCi/s.

For the purpose of implementing the controls of Section 3.1.2.b, only 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation.

All other terms were defined previously.

3.4.2 Noble Gas Air Dose at or Beyond Site Boundary For the purpose of implementing the controls of Section 3.1.3, air doses in areas at or beyond the SITE BOUNDARY due to releases of noble gases from each unit shall be calculated as follows (adapted from Reference 1, page 28, by including only long-term releases):

(3.14)

(3.15) 3-27 Version 24 01/10

where:

8 7 3.17 X 10- = a units conversion factor: 1 y/(3.15 x 10 s).

D~ = the air dose due to beta emissions from noble gas radionuclides, in mrad.

Dy = the air dose due to gamma emissions from noble gas radionuclides, in mrad.

Ni = the air dose factor due to beta emissions from noble gas radionuclide i, in (mrad/y)/(/lCi/m 3 ), from Table 3-5.

Mi = the air dose factor due to gamma emissions from noble gas radionuclide i, in (mrad/y)/(/lCi/m 3 ), from Table 3-5.

~

Qiv = the cumulative release of noble gas radionuclide i from release pathway v, in /lCi, during the period of interest.

All other terms were defined previously.

Because the air dose limit is on a per-reactor-unit basis, the summations extend over all gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned to the two units in any reasonable manner, provided that all activity released via the particular shared release pathway is apportioned to one or the other unit.

The gaseous final release pathways at the plant site, and the (X/Q)Vb for each, are identified in Table 3-4.

3-28 Version 24 01/10

FNP*ODCM Table 3*5 Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases y - Body (K) ~ - Skin (L) 'Y - Air (M) ~ - Air (N)

Nuclide (mrem/y) per (mrem/y) per (mrad/y) Rer (mrad/y) Rer (IlCi/m3) (IlCi/m3) (IlCi/m ) (IlCi/m )

Kr-83m 7.56 E-02 0.00 E+OO 1.93 E+01 2.88 E+02 Kr-85m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 Kr-85 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03 Kr-87 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Xe-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+02 Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 All values in this table were obtained from Reference 3 (Table B-1), with units converted.

3-29 Version 24 01/10

FNP-ODCM Table 3-6 Dose Factors for Exposure to Direct Radiation from Noble Gases in an Elevated Finite Plume The contents of this table are not applicable to the Farley Nuclear Plant.

3-30 Version 24 01/10

FNP*ODCM 3.4.3 Dose to a Member of the Public at or Beyond Site Boundary The dose received by an individual due to gaseous releases from each reactor unit, to areas at or beyond the SITE BOUNDARY, depends on the individual's location, age group, and exposure pathways. The MEMBER OF THE PUBLIC expected to receive the highest dose in the plant vicinity is referred to as the controlling receptor. The dosimetrically-significant attributes of the currently-defined controlling receptor are presented in Table 3-7.

Doses to a MEMBER OF THE PUBLIC due to gaseous releases of 1-131,1-133, tritium, and all radionuclides in particulate form from each unit shall be calculated as follows (equation adapted from Reference 1, page 29, by considering only long-term releases):

Dja = 3.17 x 10- 8 Z:{~RaiPj Z:[Wvip . Q iv]l (3.16) p l v 1 where:

Dja = the dose to organ j of an individual in age group a, due to gaseous releases of 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in mrem.

8 3.17 x 10- = a units conversion factor: 1 y/(3.15 x 107 s).

Raipj = the site-specific dose factor for age group a, radionuclide i, exposure pathway p, and organ j. For the purpose of implementing the controls of Section 3.1.4, the exposure pathways applicable to calculating the dose to the currently-defined controlling receptor are included in Table 3-7; values of RaiPl for each exposure pathway and radionuclide applicable to calculations of dose to the controlling receptor are listed in Table 3-8 through Table 3-11.

A detailed discussion of the methods and parameters used for calculating Raipj for the plant site is presented in Chapter 9. That information may be used for recalculating the RaiPi values if the underlying parameters change, or for calculating RaiPi values for special radionuclides and age groups when performing the assessments discussed in Section 3.4.4 below.

Wvip = the annual average relative dispersion or deposition at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radionuclide i.

For all tritium pathways, and for the inhalation of any radionuclide: Wvip is (XlQ)vp, the annual average relative dispersion factor for release pathway v, at the location of the controlling receptor (s/m \ For the ground-plane exposure pathway, and for all ingestion-related pathways for radionuclides other than tritium: Wvip is (D/Q)vp, the annual average 3-31 Version 24 01/10

FNP*ODCM relative deposition factor for release pathway v, at the location of the

-2 -- -

controlling receptor (m ). Values of (XlQ)vp and (D/Q>Vp for use in calculating the dose to the currently-defined controlling receptor are included in Table 3-7.

Qtv = the cumulative release of radionuclide i from release pathway v, during the period of interest (/lCi). For the purpose of implementing the controls of Section 3.1.4, only 1-131,1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation. In any dose assessment using the methods of this subsection, only radionuclides detectable above background in their respective samples should be included in the calculation.

Because the MEMBER OF THE PUBLIC dose limit is on a per-unit basis, the summations extend over all gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned between the two units in any reasonable manner, provided that all activity released from the plant site is apportioned to one unit or the other.

The gaseous final release pathways at the plant site, and the release elevation for each, are identified in Table 3-4.

3-32 Version 24 01/10

FNPwODCM Attributes of the Controlling Receptor The locations of MEMBERS OF THE PUBLIC in the vicinity of the plant site, and the exposure pathways associated with those locations, were determined in the Annual Land Use Census.

Dispersion and deposition values were calculated based on site meteorological data collected for the years 1971 through 1975.

Based on the Land Use Census of June 7, 1991, the current controlling receptor for the plant site is described as follows:

Sector: SW Distance: 1.2 miles Age Group: Child Exposure Pathways: Ground plane Inhalation Garden vegetation GrasslCowlMeat Dispersion Factors <XiQ}vp:

Ground-Level discharge points: 8.74 X 10.6 s/m 3 Mixed-Mode discharge points: 8.03 x 10.7 s/m 3 Deposition Factors (OO)vp:

Ground-Level discharge paints 2.64 x 10.8 m-2 Mixed-Mode discharge points: 1.05 X 10.8 m*2 This location represents the residence with the highest annual average XJQ and D/Q factors in the vicinity of the FNP. The referenced Land Use Census identified no locations where animals are maintained for milk within 5 miles of the plant site; thus, it is very unlikely that any real dairy location (which would be beyond 5 miles) would have a higher potential dose impact than the real residence location selected.

3-33 Version 24 01/10

FNP-ODCM 3.4.4 Dose Calculations to Support Other Requirements Case 1: Under 10 CFR 50.72 and 10 CFR 50.73, a radiological impact assessment may be required to support evaluation of a reportable event.

Dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the dispersion and deposition parameters [(X/Q) and (D/Q)] for the period covered by the report, and using the appropriate pathway dose factors (RaiPj ) for the receptor of interest. Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8.

The values of Raipj presented in Table 3-8 through Table 3-11 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor, Raipj values applicable to that receptor must first be calculated.

Methods and parameters for calculating Ra;pj for radionuclides and age groups other than those required in Section 3.4.3 are presented in Chapter 9. When calculating RaiPj for evaluation of an event, pathway and usage factors specific to the receptor involved in the event may be used in place of the values in Chapter 9, if the specific values are known.

Case 2: A dose calculation is required to evaluate the results of the Land Use Census, under the provisions of Section 4.1.2.

In the event that the Land Use Census reveals that exposure pathways have changed at previously-identified locations, or if new locations are identified, it may be necessary to calculate doses at two or more locations to determine which should be deSignated as the controlling receptor. Such dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the annual average dispersion and deposition values

[(X/Q) and (D/Q)] for the locations of interest, and using the appropriate pathway dose factors (Ra;pj) for the receptors of interest.

Methods for calculating (X/Q) and (D/Q) from meteorological data are presented in Chapter 8. The values of Raipj presented in Table 3-8 through Table 3-11 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor, Raipj values applicable to that receptor must first be calculated. Methods and parameters for calculating Raipj for radionuclides and age groups other than those required in Section 3.4.3 are presented in Chapter 9.

Case 3: Under Section 5.2, a dose calculation is required to support determination of total dose to a receptor of age group other than that currently defined as the controlling receptor.

Dose calculations shall be performed using the equations in Section 3.4.3, using the dispersion and deposition parameters defined in Table 3-7 for the controlling receptor, but substituting the appropriate pathway dose factors (R aipj ) for the receptor age group of interest.

The values of Raipj presented in Table 3-8 through Table 3-11 are applicable only to the currently-defined controlling receptor, so that when dose calculations must be performed for a different receptor age group, Raipj values applicable to that receptor must first be calculated. Methods and parameters for calculating Ra;pj for radionuclides and age groups other than those required in Section 3.4.3 are presented in Chapter 9.

3-34 Version 24 01/10

FNP-ODCM Table 3*8 Raipj for Ground Plane Pathway, All Age Groups Nuclide T. Body Skino~

H-3 0.00 Cr-51 4.66E+06 5.51E+06 Mn-54 1.39E+09 1.63E+09 Fe-55 0.00 0.00 Fe-59 2.73E+OB 3.21E+OB Co-5B 3.79E+OB 4.44E+OB Co-60 2.15E+10 2.53E+10 Ni-63 0.00 0.00 Zn-65 7.47E+OB B.59E+OB Rb-B6 B.99E+06 1.03E+07 Sr-B9 2.16E+04 2.51E+04 Sr-90 0.00 0.00 Y-91 1.07E+06 1.21E+06 Zr-95 2.45E+OB 2.B4E+OB Nb-95 1.37E+OB 1.61E+OB Ru-103 1.0BE+OB 1.26E+OB Ru-106 4.22E+OB 5.07E+OB Ag-110m 3.44E+09 4.01E+09 Sb-124 5.9BE+OB 6.90E+OB Sb-125 2.34E+09 2.64E+09 Sb-126 B.54E+OB 9.60E+OB Te-125m 1.55E+06 2.13E+06 Te-127m 9.16E+04 1.0BE+05 Te-129m 1.9BE+07 2.31E+07 1-131 1.72E+07 2.09E+07 1-133 2.45E+06 2.9BE+06 Cs-134 6.B6E+09 8.00E+09 Cs-136 1.51E+OB 1.71E+OB Cs-137 1.03E+10 1.20E+10 Ba-140 2.05E+07 2.35E+07 Ce-141 1.37E+07 1.54E+07 Ce-144 6.95E+07 B.04E+07 Pr-143 0.00 0.00 Nd-147 B.39E+06 1.01E+07

1. Units are m2*(mrem/yr)/(IlCi/s).

2. The values in the Total Body column also apply to the Bone, Liver, Thyroid, Kidney, Lung, and GI-LU organs.

3. This table also supports the calculations of Section 6.2.

3-35 Version 24 01/10

FNP-ODCM Table 3-9 RaiPj for Inhalation Pathway, Child Age Group Nuclide Bone Liver T.Body Thyroid Kidney L -

H-3 0.00 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 1.12E+03 Cr-51 0.00 0.00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 Mn-54 0.00 4.29E+04 9.51E+03 0.00 1.00E+04 1.58E+06 2.29E+04 Fe-55 4.74E+04 2.52E+04 7.77E+03 0.00 0.00 1.11E+05 2.87E+03 Fe-59 2.07E+04 3.34E+04 1.67E+04 0.00 0.00 1.27E+06 7.07E+04 Co-58 0.00 1.77E+03 3.16E+03 0.00 0.00 1.11 E+06 3.44E+04 Co-60 0.00 1.31E+04 2.26E+04 0.00 0.00 7.07E+06 9.62E+04 Ni-63 8.21E+05 4.63E+04 2.80E+04 0.00 0.00 2.75E+05 6.33E+03 Zn-65 4.26E+04 1.13E+05 7.03E+04 0.00 7.14E+04 9.95E+05 1.63E+04 Rb-86 0.00 1.98E+05 1.14E+05 0.00 0.00 0.00 7.99E+03 Sr-89 5.99E+05 0.00 1.72E+04 0.00 0.00 2.16E+06 1.67E+05 Sr-90 1.01E+08 0.00 6.44E+06 0.00 0.00 1.48E+07 3.43E+05 Y-91 9.14E+05 0.00 2.44E+04 0.00 0.00 2.63E+06 1.84E+05 Zr-95 1.90E+05 4.18E+04 3.70E+04 0.00 5.96E+04 2.23E+06 6.11E+04 Nb-95 2.35E+04 9.18E+03 6.55E+03 0.00 8.62E+03 6.14E+05 3.70E+04 Ru-103 2.79E+03 0.00 1.07E+03 0.00 7.03E+03 6.62E+05 4.48E+04 Ru-106 1.36E+05 0.00 1.69E+04 0.00 1.84E+05 1.43E+07 4.29E+05 Ag-110m 1.69E+04 1.14E+04 9.14E+03 0.00 2.12E+04 5.48E+06 1.00E+05 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-126 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 6.73E+03 2.33E+03 9.14E+02 1.92E+03 0.00 4.77E+05 3.38E+04 Te-127m 2.49E+04 8.55E+03 3.02E+03 6.07E+03 6.36E+04 1.48E+06 7.14E+04 Te-129m 1.92E+04 6.85E+03 3.04E+03 6.33E+03 5.03E+04 1.76E+06 1.82E+05 1-131 4.81E+04 4.81E+04 2.73E+04 1.62E+07 7.88E+04 0.00 2.84E+03 1-133 1.66E+04 2.03E+04 7.70E+03 3.85E+06 3.38E+04 0.00 5.48E+03 Cs-134 6.51E+05 1.01E+06 2.25E+05 0.00 3.30E+05 1.21 E+05 3.85E+03 Cs-136 6.51E+04 1.71E+05 1.16E+05 0.00 9.55E+04 1.45E+04 4.18E+03 Cs-137 9.07E+05 8.25E+05 1.28E+05 0.00 2.82E+05 1.04E+05 3.62E+03 Ba-140 7.40E+04 6.48E+01 4.33E+03 0.00 2.11E+01 1.74E+06 1.02E+05 Ce-141 3.92E+04 1.95E+04 2.90E+03 0.00 8.55E+03 5.44E+05 5.66E+04 Ce-144 6.77E+06 2.12E+06 3.61 E+05 0.00 1.17E+06 1.20E+07 3.89E+05 Pr-143 1.85E+04 5.55E+03 9.14E+02 0.00 3.00E+03 4.33E+05 9.73E+04 Nd-147 1.08E+04 8.73E+03 6.81E+02 0.00 4.81E+03 3.28E+05 8.21E+04 Units are (mremlyr)/(IlCi/m3) for all radionuclides.

3-36 Version 24 01/10

FNP*ODCM Table 3*10 Raipj for Cow Meat Pathway, Child Age Group.

Nuclide I Bone Liver T.Body Thyroid Kidney Lung GI-LLI H-3 0.00 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 Cr-51 0.00 0.00 B.79E+03 4.BBE+03 1.33E+03 B.91E+03 4.66E+05 Mn-54 0.00 B.01E+06 2.13E+06 0.00 2.25E+06 0.00 6.72E+06 Fe-55 4.57E+OB 2.42E+OB 7.51E+07 0.00 0.00 1.37E+OB 4.49E+07 Fe-59 3.76E+OB 6.09E+OB 3.03E+OB 0.00 0.00 1.77E+OB 6.34E+OB Co-5B 0.00 1.64E+07 5.02E+07 0.00 0.00 0.00 9.5BE+07 Co-60 0.00 6.93E+07 2.04E+OB 0.00 0.00 0.00 3.B4E+OB Ni-63 2.91E+10 1.56E+09 9.91E+OB 0.00 0.00 0.00 1.05E+OB Zn-65 3.75E+OB 1.00E+09 6.22E+OB 0.00 6.30E+OB 0.00 1.76E+OB Rb-B6 0.00 5.77E+OB 3.55E+OB 0.00 0.00 0.00 3.71E+07 Sr-B9 4.B2E+OB 0.00 1.3BE+07 0.00 0.00 0.00 1.B7E+07 Sr-90 1.04E+10 0.00 2.64E+09 0.00 0.00 0.00 1.40E+OB!

Y-91 1.BOE+06 0.00 4.B2E+04 0.00 0.00 0.00 2.40E+OB Zr-95 2.66E+06 5.B5E+05 5.21E+05 0.00 B.3BE+05 0.00 6.11E+OB Nb-95 3.10E+06 1.21E+06 B.62E+05 0.00 1.13E+06 0.00 2.23E+09 Ru-103 1.55E+OB 0.00 5.96E+07 0.00 3.90E+OB 0.00 4.01E+09 Ru-106 4.44E+09 0.00 5.54E+OB 0.00 5.99E+09 0.00 6.90E+10 Ag-110m B.39E+06 5.67E+06 4.53E+06 0.00 1.06E+07 0.00 6.74E+OB Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-126 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 5.69E+OB 1.54E+OB 7.59E+07 1.60E+OB 0.00 0.00 5.49E+OB Te-127m 1.77E+09 4.7BE+OB 2.11E+OB 4.24E+OB 5.06E+09 0.00 1.44E+09 Te-129m 1.79E+09 5.00E+OB 2.7BE+OB 5.77E+OB 5.26E+09 0.00 2.1BE+09 1-131 1.65E+07 1.66E+07 9.46E+06 5.50E+09 2.73E+07 0.00 1.4BE+06 1-133 5.67E-01 7.02E-01 2.66E-01 1.30E+02 1.17E+00 0.00 2.B3E-01 Cs-134 9.22E+OB 1.51 E+09 3.19E+OB 0.00 4.69E+OB 1.6BE+OS B.16E+06 Cs-136 1.62E+07 4.46E+07 2.BBE+07 0.00 2.37E+07 3.54E+06 1.57E+06 Cs-137 1.33E+09 1.2BE+09 1.SBE+OB 0.00 4.16E+OS 1.50E+OB 7.99E+06 Ba-140 4.3BE+07 3.B4E+04 2.56E+06 0.00 1.25E+04 2.29E+04 2.22E+07 Ce-141 2.22E+04 1.11E+04 1.64E+03 0.00 4.B6E+03 0.00 1.3BE+07 Ce-144 2.32E+06 7.26E+05 1.24E+05 0.00 4.02E+05 0.00 1.B9E+OB Pr-143 3.34E+04 1.00E+04 1.66E+03 0.00 5.43E+03 0.00 3.60E+07 Nd-147 1.17E+04 9.47E+03 7.33E+02 0.00 5.19E+03 0.00 1.50E+07 Units are (mrem/yr)/(IlCilm 3 ) for tritium, and m2.(mrem/yr)/(IlCi/s) for all other radionuclides.

3-37 Version 24 01/10

FNP-ODCM Table 3-11 RaiPj for Garden Vegetation Pathway, Child Age Group.

Nuclide Bone ~ver T.Body Thyroid Kidney Lung GI-LLI H-3 0.00 4.01E+03 4.01 E+03 4.01E+03 4.01E+03 4.01E+03 4.01 E+03 Cr-S1 0.00 0.00 1.17E+OS 6.S0E+04 1.7BE+04 1.19E+OS 6.21E+06 Mn-S4 0.00 6.6SE+OB 1.77E+OB 0.00 1.B6E+OB 0.00 S.SBE+OB Fe-SS B.01E+OB 4.2SE+OB 1.32E+OB 0.00 0.00 2.40E+OB 7.B7E+07 Fe-S9 3.9BE+OB 6.43E+OB 3.20E+OB 0.00 0.00 1.B6E+OB 6.70E+OB Co-SB 0.00 6.44E+07 1.97E+OB 0.00 0.00 0.00 3.76E+OB Co-60 0.00 3.7BE+OB 1.12E+09 0.00 0.00 0.00 2.10E+09 Ni-63 3.9SE+10 2.11E+09 1.34E+09 0.00 0.00 0.00 1.42E+OB Zn-6S B.13E+OB 2.16E+09 1.3SE+09 0.00 1.36E+09 0.00 3.BOE+OB Rb-B6 0.00 4.S2E+OB 2.7BE+OB 0.00 0.00 0.00 2.91E+07 Sr-B9 3.60E+10 0.00 1.03E+09 0.00 0.00 0.00 1.39E+09 Sr-90 1.24E+12 0.00 3.1SE+11 0.00 0.00 0.00 1.67E+10 Y-91 1.B6E+07 0.00 4.99E+OS 0.00 0.00 0.00 2.4BE+09 Zr-9S 3.B6E+06 B.4BE+OS 7.SSE+OS 0.00 1.21E+06 0.00 B.BSE+OB Nb-9S 4.10E+OS 1.60E+OS 1.14E+OS 0.00 1.S0E+OS 0.00 2.96E+OB Ru-103 1.S3E+07 0.00 S.90E+06 0.00 3.B6E+07 0.00 3.97E+OB Ru-106 7.4SE+OB 0.00 9.30E+07 0.00 1.01E+09 0.00 1.16E+10 Ag-110m 3.21E+07 2.17E+07 1.73E+07 0.00 4.04E+07 0.00 2.SBE+09 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-12S 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-126 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-12Sm 3.S1E+OB 9.S0E+07 4.67E+07 9.B4E+07 0.00 0.00 3.3BE+OB Te-127m 1.32E+09 3.S6E+OB 1.S7E+OB 3.16E+OB 3.77E+09 0.00 1.07E+09 Te-129m B.41E+OB 2.3SE+OB 1.31E+OB 2.71 E+OB 2.47E+09 0.00 1.03E+09 1-131 1.43E+OB 1.44E+OB B.17E+07 4.7SE+10 2.36E+OB 0.00 1.2BE+071 1-133 3.S3E+06 4.37E+06 1.6SE+06 B.11E+OB 7.2BE+06 0.00 1.76E+06 Cs-134 1.60E+10 2.63E+10 S.SSE+09 0.00 B.1SE+09 2.93E+09 1.42E+OB Cs-136 B.24E+07 2.27E+OB 1.47E+OB 0.00 1.21E+OB 1.BOE+07 7.96E+06 Cs-137 2.39E+10 2.29E+10 3.3BE+09 0.00 7.46E+09 2.6BE+09 1.43E+OB Ba-140 2.77E+OB 2.42E+OS 1.61E+07 0.00 7.B9E+04 1.4SE+OS 1.40E+OB Ce-141 6.S6E+OS 3.27E+OS 4.B6E+04 0.00 1.43E+OS 0.00 4.0BE+OB Ce-144 1.27E+OB 3.9BE+07 6.7BE+06 0.00 2.21E+07 0.00 1.04E+10 Pr-143 1.46E+OS 4.37E+04 7.23E+03 0.00 2.37E+04 0.00 1.S7E+OB Nd-147 7.1SE+04 S.79E+04 4.4BE+03 0.00 3.1BE+04 0.00 9.17E+07 2

Units are (mrem/yr)/{I-ICi/m 3 ) for tritium, and m .(mrem/yr)/(I-ICi/s) for all other radionuclides.

3-3B Version 24 01/10

FNP-ODCM 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS 3.5.1 Thirty-One Day Dose Projections In order to meet the requirements of the limit for operation of the gaseous radwaste treatment system (see Section 3.1.5), dose projections must be made at least once each 31 days; this applies during periods in which a discharge to areas at or beyond the SITE BOUNDARY of gaseous effluents containing radioactive materials occurs or is expected.

Projected 31-day air doses and doses to individuals due to gaseous effluents may be determined as follows:

For air doses:

D JDPc JX3J+Dpa

{Jp l t (3.17)

For individual doses:

DOP=(DOC)X31+D (3.18) t oa where:

Dpp = the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases.

Ope = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

Dpa = the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, Dpa may be set to zero.

Dyp = the projected air dose due to gamma emissions from noble gases for the next 31 days of gaseous releases.

Dye = the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

3-39 Version 24 01/10

FNP*ODCM D-ya = the anticipated air dose due to gamma emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, D-ya may be set to zero.

Dop = the projected dose to the total body or organ 0, due to releases of 1-131, 1-133, tritium, and particulates for the next 31 days of gaseous releases.

Doc = the cumulative dose to the total body or organ 0, due to releases of 1-131, 1-133, tritium, and particulates that have occurred in the elapsed portion of the current quarter, plus the release under consideration.

Doa = the anticipated dose to the total body or organ 0, due to releases of 1-131, 1-133, tritium, and particulates, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, Doa may be set to zero.

t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter).

3.5.2 Dose Projections for Specific Releases Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For air dose and individual dose projections due to gaseous effluent releases, follow the methodology of Section 3.4, using sample analysis results for the gaseous stream to be released, and parameter values expected to exist during the release period.

3-40 Version 24 01/10

FNp*ODCM 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS Section of Term Definition Initial use AG = the administrative allocation factor for gaseous streams, applied to divide the gaseous release limit among all the release pathways [unitless]. 3.3.2.1 AG s = the administrative allocation factor for gaseous source stream s, applied to divide the gaseous release limit among all the release pathways [unitless]. 3.3.3 AG v = the administrative allocation factor for gaseous release pathway v, applied to divide the gaseous release limit among all the release pathways [unitless1. 3.3.2.2 c = the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line prior to release [IlCi/mL]. 3.3.2.1 Cns = the calculated noble gas efRuent monitor setpoint for gaseous source stream s [IlCi/mL1. 3.3.3 cnv = the calculated noble gas effluent monitor setpoint for release pathway v [IlCi/mL]. 3.3.2.2 Dja = the dose to organ j of an individual in age group a, due to gaseous releases ofl-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days [mrem]. 3.4.3 Doa = the anticipated dose to organ 0 due to releases of non-noble-gas radionuclides, contributed by any planned activities during the next 31-day period [mrem1. 3.5.1 Doc = the cumulative dose to organ 0 due to releases of non-noble-gas radionuclides that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrem]. 3.5.1 Dop = the projected dose to organ 0 due to the next 31 days of gaseous releases of non-noble-gas radionuclides [mrem]. 3.5.1

= the air dose due to beta emissions from noble gas DI3 radionuclides [mrad]. 3.4.2 Dpa = the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period [mrad]. 3.5.1 3-41 Version 24 01/10

FNP-ODCM Section of Term Definition Initial use Dpe = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrad1. 3.5.1 Dpp = the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases [mrad]. 3.5.1 Dy = the air dose due to gamma emissions from noble gas radionuclides [mrad1. 3.4.2

= the anticipated air dose due to gamma emissions from noble D"<<l gas releases, contributed by any planned activities during the next 31-day period [mrad1. 3.5.1 Dye = the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrad1. 3.5.1

= the projected air dose due to gamma emissions from noble Dw gases, for the next 31 days of gaseous releases [mrad1. 3.5.1 (D/Q)vp= the annual average relative deposition factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [m-21. 3.4.3 DRk = the skin dose rate at the time of the release [mrem/y1. 3.4.1.1 DRo = the dose rate to organ 0 at the time of the release [mrem/y]. 3.4.1.2 DRt = the total body dose rate at the time of the release [mrem/y]. 3.4.1.1 fav = the maximum anticipated actual discharge flowrate for release pathway v during the period of the planned release

[mUs]. 3.3.2.2 fas = the maximum anticipated actual discharge flowrate for gaseous source stream s during the period of the planned release [mUs]. 3.3.3 Kj = the total body dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5 [(mrem/y)/(IlCi/m3)]. 3.3.2.2 Lj = the skin dose factor due to beta emissions from noble gas radionuclide i, from Table 3-5 [(mrem/y)/(IlCilm 3)J. 3.3.2.2 Mj = the air dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5 [(mrad/y)/(IlCi/m3)]. 3.4.2 3-42 Version 24 01/10

FNP*ODCM Section of Term Definition Initial use N = the number of simultaneously active gaseous release pathways [unitless]. 3.3.4 Ni = the air dose factor due to beta emissions from noble gas radionuclide i, from Table 3-5 [(mrad/y)/(!-lCi/m 3 )]. 3.4.2 Pio = the site-specific dose factor for radionuclide i (1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) and organ o. The values of PiO are equal to the site-specific Raipj values presented in Table 3-9 [(mrem/y)/{!-lCi/m 3 )]. 3.4.1.2 Oiv = the release rate of noble gas radionuclide i from release pathway v during the period of interest [!-lCi/s]. 3.3.2.2 I

Qiv= the release rate of radionuclide i (1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) from gaseous release pathway v during the period of interest [!-lCi/s]. 3.4.1.2 Qiv = the cumulative release of noble gas radionuclide i from release pathway v during the period of interest [!-lCi]. 3.4.2

- I Qiv = the cumulative release of non-noble-gas radionuclide i from release pathway v, during the period of interest [!-lCi]. 3.4.3 Raipj = the site-specific dose factor for age group a, radionuclide i, exposure pathway p, and organ j. Values and units of Raipj for each exposure pathway, age group, and radionuclide that may arise in calculations for implementing Section 3.1.4 are listed in Table 3-8 through Table 3-11. 3.4.3 Rk = the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in the release under consideration [unitless]. 3.3.2.1 Rt = the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the release under consideration [unitless]. 3.3.2.1 the ratio of the skin dose rate limit for noble gases, to the rk = skin dose rate due to noble gases in the source stream under consideration [unitless]. 3.3.3.1 3-43 Version 24 01/10

FNP-ODCM Section of Term Definition Initial use rt = the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the source stream under consideration [unitless]. 3.3.3.1 SF = the safety factor used in gaseous setpoint calculations to compensate for statistical fluctuations and errors of measurement [unitless]. 3.3.2.2 t = the number of whole or partial days elapsed in the current quarter, including the period of the release under consideration. 3.5.1 Wvip = the annual average relative dispersion [(XlQ)vp] or deposition [(D/Q)vp] at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radionuclide i.

3.4.3 X = the noble gas concentration for the release under consideration [IlCi/mL]. 3.3.2.1 the concentration of radionuclide i applicable to active Xir = gaseous release pathway r [IlCi/mLl 3.3.4 Xis = the measured concentration of radionuclide i in gaseous source stream s [IlCi/mL]. 3.3.3 the measured concentration of radionuclide i in gaseous stream v Xiv = [IlCi/mL]. 3.3.2.2 (XlQ) = the highest relative concentration at any point at or beyond the SITE BOUNDARY [s/m 3]. 3.3.2.1 the annual average SITE BOUNDARY relative concentration (XlQ)rb = applicable to active gaseous release pathway r [s/m 3].

3.3.4 (XlQ)vb = the highest annual average relative concentration at the SITE BOUNDARY for the discharge pOint of release pathway v, from Table 3~4 [s/m 3]. 3.3.2.2 (XiU)vp = annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, from Table 3~7 [s/m 3]. 3.4.3 344 Version 24 01/10

FNP-ODCM CHAPTER 4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.1 LIMITS OF OPERATION The following limits are the same for both units at the site. Thus, a single program including monitoring, land use survey, and quality assurance serves both units.

4.1.1 Radiological Environmental Monitoring In accordance with this ODCM, the Radiological Environmental Monitoring Program (REMP) shall be conducted as specified in Table 4-1.

4.1.1.1 Applicability This control applies at all times.

4.1.1.2 Actions 4.1.1.2.1 With the REMP not being conducted as specified in Table 4-1, submit to the Nuclear Regulatory Commission (NRC), in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations from the required sampling schedule are permitted if specimens are unobtainable due to hazardous conditions, unavailability, inclement weather, equipment malfunction, or other just reasons. If deviations are due to equipment malfunction, efforts shall be made to complete corrective action prior to the end of the next sampling period.

4.1.1.2.2 With the confirmed 1 measured level of radioactivity as a result of plant effluents in an environmental sampling medium specified in Table 4-1 exceeding the reporting levels of Table 4-2 when averaged over any calendar quarter, submit within 30 days a Special Report to the NRC pursuant to 10 CFR 50.4. The Special Report shall identify the cause(s) for exceeding the limit(s) and define the corrective action(s) to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Sections 2.1.3, 3.1.3, and 3.1.4. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in the Special Report.

When more than one of the radionuclides in Table 4-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) + ... ;: -: 1.0 reporting level (1) reporting level (2)

When radionuclides other than those in Table 4-2 are detected and are the result of plant effluents, this Special Report shall be submitted if the potential annual dose to a MEMBER OF Defined as confirmed by reanalysis of the original sample, or analysis of a duplicate or new sample, as appropriate. The results of the confirmatory analysis shall be completed at the earliest time consistent with the analysis.

4-1 Version 24 01/10

FNP*ODCM THE PUBLIC is equal to or greater than the calendar year limits stated in Sections 2.1.3, 3.1.3, and 3.1.4. This Special 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 described in the Annual Radiological Environmental Operating Report. The levels of naturally-occurring radionuclides which are not included in the plant's effluent releases need not be reported.

4.1.1.2.3 If adequate samples of milk, or during the growing season, forage or fresh leafy vegetation, can no longer be obtained from one or more of the sample locations required by Table 4-1, or if the availability is frequently or persistently wanting, efforts shall be made: to identify specific locations for obtaining suitable replacement samples; and to add any replacement locations to the REMP given in the ODCM within 30 days. The specific locations from which samples became unavailable may be deleted from the REMP. Pursuant to Technical Specification 5.5.1, documentation shall be submitted in the next Radioactive Effluent Release Report for the change(s) in the ODCM, including revised figure(s) and table(s) reflecting the changes to the location(s), with supporting information identifying the cause of the unavailability of samples and justifying the selection of any new location(s).

4.1.1.2.4 This control does not affect shutdown requirements or MODE changes.

4.1.1.3 Surveillance Requirements The REMP samples shall be collected pursuant to Table 4-1 from the locations described in Section 4-2, and shall be analyzed pursuant to the requirements of Table 4-1 and Table 4-3.

Program changes may be initiated based on operational experience.

Analyses shall be performed in such a manner that the stated MINIMUM DETECTABLE CONCENTRATIONs (MDCs) will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering radionuclides, or other uncontrollable circumstances may render these MDCs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

4.1.1.4 Basis The REMP required by this control provides representative measurements of radiation and of radioactive materials in those exposure pathways, and for those radionuclides, which lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation. The REMP implementsSection IV.B.2, Appendix 1,10 CFR 50, and thereby supplements the radiological effluent monitoring program by measuring concentrations of radioactive materials and levels of radiation, which may then be compared with those expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.

The detection capabilities required by Table 4-3 are within state-of-the-art for routine environmental measurements in industrial laboratories.

Version 24 01/10

FNP*ODCM Table 4*1 Radiological Environmental MonitorinjJ Pr()gram Number of Exposure Pathway Samples and Sampling and Collection and/or Sample Sample FREQUENCY Type and Frequency of Analysis Locations"

1. AIRBORNE Particulates Continuous operation of sampler Particulate sampler. Analyze for gross beta Indicator 3 with sample collection weekly. radioactivity ;z: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following filter change.

Control 2 Perform gamma isotopic analysis on each sample when gross beta activity is > 10 times the yearly mean of control samples. Perform gamma isotopic analysis on composite (by location) sample quarterly.

Radioiodine Indicator 3 Control 2 Radioiodine canister. Analyze weekly for 1-131.

2. DIRECT RADIATION TLD Quarterly. Gamma dose quarterly.

Indicator I 16 Indicator II 16 (community)

Control 3

3. WATERBORNE Surface Composite"" sample collected Gamma isotopic analyses monthly. Tritium analysis Indicator 1 monthly. of composite (by location) sample quarterly.

Control 1 Offsite Ground Gamma isotopic and tritium analysis of each Indicator 1 Quarterly. sample.

Control 1 Sediment Gamma isotopic analysis of each sample.

Semiannually.

Indicator 1 4-3 Version 24 01/10

FNP*QDCM Table 4-1 (contd) Radiological Envirclnmel1tal MonitoringErogram Onsite Ground Tritium and gamma isotopic quarterly.

18 Indicator Quarterly. Other analyses based on results of 3

Control tritium and gamma.

Number of Exposure Pathway Samples and and/or Sample Sampling and Collection Frequency Type and Frequency of Analysis Sample Locations*

4. INGESTION Milk Semimonthly when animals are on pasture; Gamma isotopic and 1-131 analysis of Indicator 3*** monthly at other times. each sample.

Control 1 One sample in season, or semiannually if Gamma isotopic analysis on edible Fish 1 not seasonal. One sample of each of the portions.

Indicator 1 following species:

Control 1. Game Fish

2. Bottom Feeding Fish Forage or Grab sample cut from green forage or Gamma isotopic analysis which Leafy 1 vegetation monthly. includes 1-131 analysis of each sample.

Vegetation 1 Indicator Control

Sample locations are shown in Table 4-4 and Table 4-5 and in Figure 4-1 through Figure 4-5.

- Composite samples shall be collected by collecting an aliquot at intervals not exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

- - Up to three sampling locations within 5 miles in different sectors with the highest dose potential will be used as available.

4-4 Version 24 01/10

FNp*ODCM Table 4*2 Reporting Levels for Radioactivity Concentrations in Environmental Samples

--.-~- -- --

Reporting Level Airborne Particulate or Forage or Leafy Water Gases Fish Milk Vegetation Analysis (pCi/L) (pCi/m3) (pCilkg, wet) (pCi/L) (pCi/kg, wet)

H-3 2 E+4a Mn-54 1 E+3 3 E+4 Fe-59 4 E+2 1 E+4 Co-58 1 E+3 3 E+4 Co-60 3 E+2 1 E+4 Zn-65 3 E+2 2 E+4 Zr-95 4 E+2 Nb-95 7 E+2 1-131 2 E+Ob 9 E-1 3 E+O 1 E+2 Cs-134 3 E+1 1 E+1 1 E+3 6 E+1 1 E+3 Cs-137 5 E+1 2 E+1 2 E+3 7 E+1 2 E+3 Ba-140 2 E+2 3 E+2 La-140 1 E+2 4 E+2

a. This is the 40 CFR 141 value for drinking water samples. If no drinking water pathway exists, a value of 3 E+4 pCi/L may be used.

b. If no drinking water pathway exists, a value of 20 pCilL may be used.

4-5 Version 24 01/10

FNP*ODCM Table 4-3 Values for the Minimum Detectable Concentration I

Minimum Detectable Concentration (MDc)a i Airborne Grass or Leafy Water Particulate or Fish Milk Vegetation Sediment Analysis (pCi/L) Gases (pCi/m3) (pCi/kg, wet) (pCi/L) (pCi/kg, wet) (pCi/kg, dry) gross beta 4 E+O 1 E-2 H-3 2 E+3 D Mn-54 1.5 E+1 1.3 E+2 Fe-59 3 E+1 2.6 E+2 Co-58, Co-60 1.5 E+1 1.3 E+2 Zn-65 3 E+1 2.6 E+2 Zr-95 3 E+1 Nb-95 1.5 E+1 1-131 1 E+Oc 7 E-2 1 E+O 6 E+1 Cs-134 1.5 E+1 5 E-2 1.3 E+2 1.5 E+1 6 E+1 1.5 E+2 Cs-137 1.8 E+1 6 E-2 1.5 E+2 1.8 E+1 8 E+1 1.8 E+2 Ba-140 6 E+1 6 E+1 La-140 1.5 E+1 1.5 E+1

a. See the definition of MINIMUM DETECTABLE CONCENTRATION in Section 10.1. Other peaks which are measurable and identifiable as plant effluents, together with the radionuclides in this table, shall be analyzed and reported in accordance with Section 7.1.

b. If no drinking water pathway exists, a value of 3 E+3 pCi/L may be used.

c. If no drinking water pathway exists, a value of 1.5 E+1 pCi/L may be used.

4-6 Version 24 01110

FNp*ODCM 4.1.2 Land Use Census In accordance with this ODCM, a land use census shall be conducted and shall identify the 1

location of the nearest milk animal and the nearest permanent residence, in each of the 16 meteorological sectors, within a distance of 5 miles.

4.1.2.1 Applicability This control applies at all times.

4.1.2.2 Actions 4.1.2.2.1 With a land use census identifying a location(s) which yields a calculated dose or dose commitment greater than values currently being calculated in accordance with Section 3.4.3, identify the new location(s) in the next Radioactive Effluent Release Report.

4.1.2.2.2 With a land use census identifying a location(s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20 percent greater than at a location from which samples are currently being obtained in accordance with Section 4.1.1, add the new location(s) to the REMP within 30 days if samples are available. The sampling location, excluding control station location(s), having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from the REMP if new sampling locations are added.

Pursuant to Technical Specification 5.5.1 submit in the next Radioactive Effluent Release Report any change(s) in the ODCM, including the revised figure(s) and table(s) reflecting any new location(s} and information supporting the change(s).

4.1.2.2.3 This control does not affect shutdown requirements or MODE changes.

4.1.2.3 Surveillance Requirements The land use census shall be conducted annually, using that information which will provide good results, such as a door-to-door census, a visual census from automobile or aircraft, consultation with local agriculture authorities, or some combination of these methods, as feasible. Results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

4.1.2.4 Basis This control is provided to ensure that changes in the use of UNRESTRICTED AREAS are identified and that modifications to the REMP are made if required by the results of this census.

This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50.

lDefined as a cow or goat that is producing milk for human consumption.

4-7 Version 24 01/10

FNP*ODCM 4.1.3 Interlaboratory Comparison Program In accordance with this ODCM, analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which satisfies the requirements of Regulatory Guide 4.15, Revision 1, February 1979.

4.1.3.1 Applicability This control applies at all times.

4.1.3.2 Actions With analyses not being performed as required by Section 4.1.3, report the corrective actions taken to prevent a recurrence in the Annual Radiological Environmental Operating Report.

This control does not affect shutdown requirements or MODE changes.

4.1.3.3 Surveillance Requirements A summary of the results obtained as part of the required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report.

4.1.3.4 Basis The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring, in order to demonstrate that the results are reasonably valid for the purposes of Section IV.B.2, Appendix I, 10 CFR 50.

4*8 Version 24 01/10

FNP-ODCM 4.2 RADIOLOGICAL ENVIRONMENTAL MONITORING LOCATIONS Table 4-4, and Figure 4-1 through Figure 4-4 specify the locations at which the measurements and samples are taken for the REMP required by Section 4.1.1.

4-9 Version 24 01/10

FNP*ODCM Table 4-4 Radiological Environmental Monitoring Locations Exposure Pathway and/or Sampling Locations

. Sample Identification Sample

1. AIRBORNE Particu lates Indicator Stations:

River Intake Structure (ESE-O.B miles)1 PI-0501 South Perimeter (SSE-1.0 miles) PI-0701 Plant Entrance (WSW-0.9 miles) PI-1101 North Perimeter(N-O.B miles) PI-1601 Control Stations:

Blakely, GA (NE-15 miles) PB-0215 Dothan, AL (W-1B miles) PB-121B Neals Landing, FL (SSE-18 miles)1 PB-071B Communitv Stations:

Georgia Pacific Paper Co. (SSE-3 miles) PC-0703 Ashford, AL (WSW-8 miles) PC-110B Columbia, ALIN-5 miles) PC-1605 Radioiodine Indicator Stations:

River Intake Structure (ESE-0.8 miles)1 11-0501 South Perimeter (SSE-1.0 miles) 11-0701 Plant Entrance (WSW-0.9 miles) 11-1101 North Perimeter (N-O.B miles) 11-1601 Control Stations:

Blakely, GA (NE-15 miles) IB-0215 Dothan, AL (W-18 miles) IB-1218 Neals Landing, FL (SSE-18 miles)1 IB-0718 I Community Stations:

Georgia Pacific Paper Co. (SSE-3 miles)2 IIC-0703

2. DIRECT RADIATION TLD Indicator I Stations:

Plant Perimeter (NNE-0.9 miles) RI-0101 (NE-1.0 miles) RI-0201 (ENE-D.9 miles) RI-0301 (E-O.B miles) RI-0401 (ESE-O.B miles) RI-0501 (SE-1.1 miles) RI-0601 (SSE-1.0 miles) RI-0701 (S-1.0 miles) RI-OB01 (SSW-1.0 miles) RI-0901 (SW-O.9 miles) RI-1001 (WSW-O.9 miles) RI-1101 (W-O.B miles) RI-1201 (WNW-O.B miles) RI-1301 (NW-1.1 miles) RI-1401 (NNW-O.9 miles) RI-1501 (N-O.B miles) RI-1601 4-10 Version 24 01/10

FNP-ODCM Table 4-4 (contd) Radiological Environmental Monitoring Locations Exposure Pathway and/or

Sample Sample Sampling Locations Identification TLD (contd) Control Stations:

Blakely, GA (NE-15 miles) RB-0215 Neals Landing, FL (SSE-18 miles) RB-0718 Dothan, AL (W-15 miles) RB-1215 Dothan, AL (W-18 miles) RB-1218 Webb, AL (WNW-11 miles) RB-1311 Haleburg, AL (N-12 miles) RB-1612 Indicator II (Community) Stations:

(NNE-4 miles) RC-0104 (NE-4 miles) RC-0204 (ENE-4 miles) RC-0304 (E-5 miles) RC-0405 (ESE-5 miles) RC-0505 (SE-5 miles) RC-0605 (SSE-3 miles) RC-0703 (S-5 miles) RC-0805 (SSW-4 miles) RC-0904 (SW-1.2 miles) RC-1001 (SW-5 miles) RC-1005 (WSW-4 miles) RC-1104 (WSW-8 miles) RC-1108 (W-4 miles) RC-1204 (WNW-4 miles) RC-1304 (NW-4 miles) RC-1404 (NNW-4 miles) RC-1504 (N-5 miles) RC-1605

3. WATERBORNE Surface Indicator Station:

Georgia Pacific Paper Co. Intake Structure WRI JRiver Mile - 40)

Control Station:

Andrew Lock & Dam Upper Pier (River Mile - 47) WRB Ground Indicator Station:

Georgia Pacific Paper Co. Well (SSE-4 miles) WGI-07 Control Station:

Whatley Well (SW-1.2 miles) WGB-10 Sediment Indicator Station:

Smith's Bend {River Mile - 41 )3 RSI Control Station:

Andrews Lock & Dam Reservoir (River Mile - 48)3 RSB 4-11 Version 24 01/10

FNP*ODCM Table 4-4 (contd) Radiological Environmental Monitoring Locations Exposure Pathway Sample and/or Sample Sampling Locations Identification

4. INGESTION Milk Indicator Station:

None (There are no milk animals within 5 miles per the current land use survey)

Control Station:

None Fish Indicator Station:

Smith Bend (River Mile - 41 t Game Fish FGI Bottom Feeding Fish FBI Control Station:

Andrews Lock & Dam Reservoir (River Mile - 48t Game Fish FGB Bottom Feeding Fish FBB Forage or Indicator Stations:

Leafy South Southeast Perimeter (SSE-1.0 miles) FI-0701 Vegetation North Perimeter (N-0.8 miles) FI-1601 South Perimeter (S-1.0 miles)5 FI-0801 Northeast Perimeter (NE-1.0 miles)5 FI-0201 Control Station:

Dothan, AL (W-18 miles) FB-1218

Distance and direction as measured from the centerpoint between Unit 1 and Unit 2 plant vent stacks.

1. Not required by Section 4.1.1. Used as a spare station.

2. Not required by Section 4.1.1. Use for comparison purposes with State of GA EPD.

3. These collections are normally made at river mile 41.3 for the indicator station and mile 47.8 for the control station; however, due to river bottom sediment shifting caused by high flows, dredging, etc., collections may be made from river mile 40 to 42 for the indicator station and from river mile 47 to 49 for the control station.

4. Since a few miles of river water may be needed to obtain adequate fish samples, these river mile positions represent the approximate locations about which the catches are taken. Collections for the indicator station should be from river mile 37.5 to 42.5 and for the control station from river mile 47 to 52. (CAR 2283)

5. Alternate forage plots.

4-12 Version 24 01/10

FNP-ODCM Table 4-5 Onsite Groundwater Monitoring Locations Locationl Sample Point Aquifer Monitoring Purpose Coordinates I Monitoring Well Major Shallow aquifer N 31°13.471' Dilution line R1 W 85°06.705' Monitoring Well Major Shallow aquifer N 31°13.470' Dilution line R2 W 85°06.645' Monitoring Well Major Shallow aquifer N 31°13.410' Unit 2 RWST R3 W 85°06.627' Monitoring Well Major Shallow aquifer N 31°13.363' Unit 1 RWST R4 W 85°06.628' Monitoring Well Major Shallow aquifer N 31°13.343' Dilution line !

R5 W 85°06.216' Monitoring Well Major Shallow aquifer N 31°13.278' Dilution line I R6 W 85°06.574' Monitoring Well Major Shallow aquifer N 31°13.255' Dilution line R7 W 85°06.438' Monitoring Well Major Shallow aquifer N 31°13.291' Dilution line I

R8 W 85°06.329' Monitoring Well Major Shallow aquifer N 31°13.279' Dilution line R9 W 85°06.097' Monitoring Well Major Shallow aquifer N 31 °13.139' Dilution line R10 W 85°05.996' Monitoring Well Major Shallow aquifer N 31°13.595' Background 1 R11 W 85°07.002' Monitoring Well Major Shallow aquifer N 31 °12.873' Dilution line R13 W 85°05.944' Monitoring Well Major Shallow aquifer N 31°13.526' Background R14 W 85°06.427' PW#2 N 31°13.945' Production Well #2 supply W 85°06.557' PW#3 N 31°13.012' Production Well #3 supply W 85°06.837' CW#1 N 31°13.568' Construction Well West supply W 85°07.041' CW#2 N 31°13.574' Construction Well East supply W 85°06.845' I

FRW N 31°12.746' Firing Range Well supply W 85°06.648' I SW-1 N/A N/A Background 3 (Service Water Pond)

East YD N/A N 31°13.444' Plant outfall W 85°06.224' (East Yard Drain)

I SEYD N/A N 31°13.119' Plant outfall W 85°06.139' (Southeast Yard Drain) 4-13 Version 24 01/10

FNP*ODCM Figure 4*1 Airborne Sampling Locations, 0-5000 feet 4-14 Version 24 01/10

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FNP*ODCM 10 MILES HOUSTON COUNTY ALABAMA INDICATOR STATIONS CONTROL STATIONS

SURFACE & GROUNDWATER SURFACE WATER, FISH, & SEDIMENT

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FNP-ODCM CHAPTER 5 TOTAL DOSE DETERMINATIONS 5.1 LIMIT OF OPERATION In accordance with Technical Specification 5.5.4.j, the dose or dose commitment to any MEMBER OF THE PUBLIC over a calendar year, due to releases of radioactivity and to radiation from uranium fuel cycle sources, shall be limited to less than or equal to 25 mrem to the total body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem.
5.1.1 Applicability This limit applies at all times.
5.1.2 Actions With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Section 2.1.3, 3.1.3, or 3.1.4, calculations shall be made according to Section 5.2 methods to determine whether the limits of Section 5.1 have been exceeded. If these limits have been exceeded, prepare and submit a Special Report to the Nuclear Regulatory Commission, pursuant to 10 CFR 50.4, within 30 days, which defines the corrective actions to be taken to reduce subsequent releases to prevent recurrence of exceeding the limits of Section 5.1 and includes the schedule for achieving conformance with the limits of Section 5.1. This Special Report, as defined in 10 CFR 20.2203, shall also include an analysis which estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources (including all effluent pathways and direct radiation) for the calendar year that includes the release(s) covered by this report. This Special Report shall also describe the levels of radiation and concentrations of radioactive material involved, and the cause of the exposure levels or concentrations. If the estimated dose(s) exceeds the limits of Section 5.1, and if the release condition resulting in violation of the provisions of 40 CFR 190 has not already been corrected, the Special Report shall include a request for variance in accordance with the provisions of 40 CFR 190 and including the specified information of 40 CFR 190.11 (b). Submittal of the report is considered a timely request, and a variance is granted until staff action on the request is complete. The variance only relates to the limits of 40 CFR 190, and does not apply in any way to the requirements for dose limitation of 10 CFR Part 20, as addressed in other sections of this ODCM.
5-1 Version 24 01/10
FNP-ODCM This control does not affect shutdown requirements or MODE changes.
5.1.3 Surveillance Requirements Cumulative dose contributions from liquid and gaseous effluents and from direct radiation shall be determined in accordance with Section 5.2. This requirement is applicable only under the conditions set forth above in Section 5.1.2.
5.1.4 Basis This control is provided to meet the dose limitations of 40 CFR 190. The control requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents combined with doses due to direct radiation from the plant exceed the limits of 40 CFR 190. For sites containing up to 4 reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will describe a course of action which should result in the limitation of dose to a MEMBER OF THE PUBLIC for a calendar year to within the 40 CFR 190 limits. For the purposes of the Special Report. it may be assumed that the dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible with the exception that dose contributions from other uranium fuel cycle facilities at the same site or within a radius of 5 miles must be considered. If the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements of 40 CFR 190, the Special Report with a request for a variance (provided the release conditions resulting in violation of 40 CFR 190 have not already been corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR 190 until NRC staff action is completed. An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation which is part of the nuclear fuel cycle.
5-2 Version 24 01/10
FNP*ODCM 5.2 DEMONSTRATION OF COMPLIANCE There are no other uranium fuel cycle facilities within 5 miles of the plant site. Therefore, for the purpose of demonstrating compliance with the limits of Section 5.1, the total dose to a MEMBER OF THE PUBLIC in the vicinity of the plant site due to uranium fuel cycle sources shall be determined as follows:
(5.1) where:
Dn. = the total dose or dose commitment to the total body or organ k, in mrem.
DL = the dose to the same organ due to radioactivity discharged from the plant site in liquid effluents, calculated in accordance with Section 2.4.1, in mrem.
DG = the dose to the same organ due to non-noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated for the controlling receptor in accordance with Section 3.4.3, in mrem.
DD = the direct radiation dose to the whole body of an individual at the controlling receptor location, due to radioactive materials retained within the plant site, in mrem. Values of direct radiation dose may be determined by measurement, calculation, or a combination of the two.
DN = the external whole body dose to an individual at the controlling receptor location, due to gamma ray emissions from noble gas radionuclides discharged from the plant site in gaseous effluents, in mrem. DN is calculated as follows (equation adapted from Reference 1, page 22, by recasting in cumulative dose form):
(5.2) where:
7 3.17x10*8 = a units conversion factor: 1 y/(3.15 x 10 s).
(!iv = the cumulative release of noble gas radionuclide i from release pathway v
(!lCi), during the period of interest.
= the total-body dose factor due to gamma emissions from noble gas radionuclide i (mrem/Y)/(!lCilm 3 ), from Table 3-5.
(XJQ)vp = annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [slm 3 ].
As defined above, DL and DG are for different age groups, while DD and DN are not age group specific. When a more precise determination of Dn. is desired, values of DL and DG may be calculated for all four age groups, and those values used in equation (5.1) to determine age group specific values of D Tk ; the largest value of Dn. for any age group may then be compared to the limits of Section 5.1.
Version 24 01/10
FNP-ODCM CHAPTER 6 POTENTIAL DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY 6.1 REQUIREMENT FOR CALCULATION Current FNP effluent controls as established by this ODCM do not require assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 10-1). However, when such an assessment is desired, it should be performed in accordance with Section 6.2.
6.2 CALCULATIONAL METHOD For the purpose of performing the calculations required in Section 6.1, the dose to a member of the public inside the SITE BOUNDARY shall be determined at the locations, and for the receptor age groups, defined in Table 6-1. The dose to such a receptor at anyone of the defined locations shall be determined as follows:
(6.1) where:
Dlk :::: the total dose to the total body or organ k, in mrem.
DA :::: the dose to the same organ due to inhalation of non-noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated in accordance with Section 3.4.3, in mrem. The (XJQ) value to be used is given for each receptor location in Table 6-1; depleted (XJQ) values may be used in calculations for non-noble-gas radionuclides.
Ds :::: the dose to the same organ due to ground plane deposition of non-noble-gas radionuclides discharged from the plant site in gaseous effluents, calculated in accordance with Section 3.4.3, in mrem. The (D/Q) value to be used is given for each receptor location in Table 6-1.
Dp = the external whole body dose due to gamma ray emissions from noble gas radionuclides discharged from the plant site in gaseous effluents, calculated using equation (5.2), in mrem. The (XJQ) values that are to be used are given for each receptor location in Table 6-1.
Fo :::: the occupancy factor for the given location, which is the fraction of the year that one individual MEMBER OF THE PUBLIC is assumed to be present at the receptor location [unitless]. Values of Fa for each receptor location are included in Table 6-1.
6-1 Version 24 01/10
FNP-ODCM Table 6*1 Attributes of MEMBER OF THE PUBLIC Receptor Locations Inside the Site Boundary Location: Visitor Center, WSW at 0.19 miles Age Group: Child Occupancy Factor: 1.37 E-03 (based on 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per year)
Dispersion and Deposition Parameters:
Parameter Ground-Level Mixed-Mode 3
(X/O). s/m 1.04 E-04 8.80 E-06 (D/O), m-2 4.80 E-07 6.20 E-08 Location: Service Water Pond, SSW at 0.60 miles Age Group: Child Occupancy Factor: 7.57 E-03 (based on 66 hours7.638889e-4 days 0.0183 hours 1.09127e-4 weeks 2.5113e-5 months per year)
Dispersion and Deposition Parameters:
Parameter I Ground-Level Mi a.
(X/O), s/m 3 4.74 E-05 9.75 E-07 (D/O), m-2 1.31 E-07 2.78 E-08 6-2 Version 24 01/10
FNP-ODCM Table 6-1 (contd) Attributes of MEMBER OF THE PUBLIC Receptor Locations Inside the Site Boundary Location: River Water Discharge, SE at 1.02 miles Age Group: Child Occupancy Factor: 1.14 E-02 (based on 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months per year)
Dispersion and Deposition Parameters:
Parameter 1.63 E-05 7.05 E-07 4.55 E-08 1.39 E-08 6-3 Version 24 01/10
FNP-ODCM CHAPTER 7 REPORTS 7.1 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 7.1.1 Reg uirement for Re port In accordance with Technical Specification 5.6.2 and 5.6.3, the Annual Radiological Environmental Operating Report covering the REMP activities during the previous calendar year shall be submitted before May 15 of each year. (A single report fulfills the requirements for both units.) The material provided shall be consistent with the objectives outlined in Section 4.1 and Section 7.1.2 of the ODCM, and in Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.
7.1.2 Report Contents The materials specified in the following subsections shall be included in each Annual Radiological Environmental Operating Report:
7.1.2.1 Data The report shall include summarized and tabulated results of all REMP samples required by Table 4-1 taken during the report period, in a format similar to that contained in Table 3 of the Radiological Assessment Branch Technical Position (Reference 13); the results for any additional samples shall also be included. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results; the missing data shall be submitted as soon as possible in a supplementary report. The results for naturally-occurring radionuclides not included in plant effluents need not be reported.
7.1.2.2 Evaluations Interpretations and analyses of trends of the results shall be included in the report, including the following: (as appropriate) comparisons with pre-operational studies, operational controls, and previous environmental operating reports; and an assessment of any observed impacts of the plant operation on the environment. If the measured level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 4-2 is not the result of plant effluents, the condition shall be described as required by Section 4.1.1.2.2.
7.1.2.3 Programmatic Information Also to be included in each report are the following: a summary description of the REMP; a map(s) of all sampling locations keyed to a table giving distances and directions from the center point between the Unit 1 and Unit 2 plant vent stacks; the results of land use censuses required by Section 4.1.2; and the results of licensee participation in the Interlaboratory Comparison Program required by Section 4.1.3.
7-1 Version 24 01/10
FNP-ODCM 7.1.2.4 Descriptions of Program Deviations Discussions of deviations from the established program must be included in each report, as follows:
7.1.2.4.1 If the REMP is not conducted as required in Table 4-1, a description of the reasons for not conducting the program as required, and the plans for preventing a recurrence, must be included in the report.
7.1.2.4.2 If the MDCs required by Table 4-3 are not achieved, the contributing factors must be identified and described in the report.
7.1.2.4.3 If Interlaboratory Comparison Program analyses are not performed as required by Section 4.1.3, the corrective actions taken to prevent a recurrence must be included in the report.
7-2 Version 24 01/10
FNP*ODCM 7.2 RADIOACTIVE EFFLUENT RELEASE REPORT 7.2.1 Requirement for Report In accordance with Technical Specification 5.6.2 and 5.6.3, the Radioactive Effluent Release Report covering the operation of the units during the previous calendar year of operation shall be submitted before May 1 of each year. (A single submittal may be made for Units 1 and 2.
However, the submittal shall specify the releases of radioactive material in liquid and gaseous effluents from each unit and solid radioactive waste from the site.) The report shall include a summary ofthe quantities of radioactive liquid and gaseous effluents and solid waste released from the units. The material provided shall be consistent with the objectives outlined throughout this ODCM and the Process Control Program (PCP) and in conformance with 10 CFR Part 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.
7.2.2 Report Contents The materials specified in the following subsections shall be included in each Radioactive Effluent Release Report:
7.2.2.1 Quantities of Radioactive Materials Released The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the units as outlined in NRC Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with liquid and gaseous effluent data summarized on a quarterly basis and solid radioactive waste data summarized on a semiannual basis following the format of Appendix B thereof.
Unplanned releases of radioactive materials in gaseous and liquid effluents from the site to UNRESTRICTED AREAS shall be included in the report, tabulated either by quarter or by event.
For gamma emitters released in liquid and gaseous effluents, in addition to the prinCipal gamma emitters for which MDCs are specifically established in Table 2-3 and Table 3-3, other peaks which are measurable and identifiable also shall be identified and reported.
7.2.2.2 Meteorological Data The report shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing of wind speed, wind direction, and atmospheric stability, and precipitation (if measured) on magnetic tape; or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. In lieu of submission with the Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.
7.2.2.3 Dose Assessments The report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from each unit during the previous calendar year. Historical annual average meteorology or the meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents (as determined by sampling frequency and measurement) shall be used for determining the gaseous pathway dose. This assessment of radiation doses shall be performed in accordance with Sections 2.1.3, 2.4, 3.1.3, 3.1.4, 3.4.2, 3.4.3, 5.1, and 5.2.
7-3 Version 24 01/10
FNP*ODCM If a determination is required by Section 5.1.2, the report shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation; this dose assessment must be performed in accordance with Chapter 5.
7.2.2.4 Solid Radwaste Data For each type of solid waste shipped offsite during the report period, the following information shall be included:
a. Container volume,
b. Total curie quantity (specify whether determined by measurement or estimate),
c. Principal radionuclides (specify whether determined by measurement or estimate),
d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),
e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and
f. Solidification agent (e.g., cement, urea formaldehyde).
7.2.2.5 Licensee Initiated Document Changes Licensee initiated changes shall be submitted to the Nuclear Regulatory Commission as a part of or concurrent with the Radioactive Effluent Release Report for the period in which any changes were made. Such changes to the ODCM shall be submitted pursuant to Technical Specification 5.5.1. This requirement includes:
7.2.2.5.1 Any changes to the sampling locations in the radiological environmental monitoring program, including any changes made pursuant to Section 4.1.1.2.3. Documentation of changes made pursuant to Section 4.1.1.2.3 shall include supporting information identifying the cause of the unavailability of samples.
7.2.2.5.2 Any changes to dose calculation locations or pathways, including any changes made pursuant to Section 4.1.2.2.2.
7.2.2.6 Descriptions of Program Deviations Discussions of deviations from the established program shall be included in each report, as follows:
7.2.2.6.1 The report shall include deviations from composite sampling requirements included in Table 2-3 and Table 3-3.
7.2.2.6.2 The report shall include deviations from Minimum Detectable Concentration (MDC) requirements included in Table 2-3 and Table 3-3.
7.2.2.6.3 The report shall include deviations from the liquid and gaseous effluent monitoring instrumentation OPERABILITY requirements included in Sections 2.1.1 and 3.1.1, 7-4 Version 24 01/10
FNP-ODCM respectively. The report shall include an explanation as to why the inoperability of the liquid or gaseous effluent monitoring instrumentation was not corrected within the specified time requirement.
7.2.2.7 Major Changes to Radioactive Waste Treatment Systems As required by Sections 2.1.5 and 3.1.6, licensee initiated MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (liquid and gaseous) shall be reported to the Nuclear Regulatory Commission in the Radioactive Effluent Release Report covering 1
the period in which the change was reviewed and accepted for implementation.
The discussion of each change shall contain:
7.2.2.7.1 A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59; 7.2.2.7.2 Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; 7.2.2.7.3 A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems; 7.2.2.7.4 An evaluation of the change, which shows the predicted releases of rad ioactive materials in liquid and gaseous effluents that differ from those previously predicted in the license application and amendments thereto; 7.2.2.7.5 An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; 7.2.2.7.6 A comparison ofthe predicted releases of radioactive materials, in liquid and gaseous effluents, to the actual releases for the period prior to when the changes are to be made; 7.2.2.7.7 An estimate of the exposure to plant operating personnel as a result of the change; and 7.2.2.7.8 Documentation of the fact that the change was reviewed and found acceptable by the PORCo In lieu of inclusion in the Radioactive Effluent Release Report, this same information may be submitted as part of the annual FSAR update.
7-5 Version 24 01/10
FNP*ODCM 7.3 MONTHLY OPERATING REPORT This ODCM establishes no requirements pertaining to the Monthly Operating Report.
7.4 SPECIAL REPORTS Special reports shall be submitted to the Nuclear Regulatory Commission in accordance with 10 CFR 50.4, as required by Sections 2.1.3.2, 2.1.4.2, 3.1.3.2, 3.1.4.2, 3.1.5.2, 4.1.1.2.2, and 5.1.2.
7-6 Version 24 01/10
FNP-ODCM CHAPTER 8 METEOROLOGICAL MODELS The models presented in this chapter are those which were used to compute the specific values of meteorology-related parameters that are referenced throughout this ODCM. These models should also be used whenever it is necessary to calculate values of these parameters for new locations of interest.
Note: Although Plant Farley has no pure elevated releases, the sections on elevated-mode calculations (8.1.2 and 8.2.2) are included for convenience in calculating mixed-mode values. and to preserve section number compatibility with the ODCMs of the other plants in the Southern Nuclear system.
8.1 ATMOSPHERIC DISPERSION Atmospheric dispersion may be calculated using the appropriate form of the sector-averaged Gaussian model. Gaseous release elevations may be considered to be either at ground-level.
elevated, or mixed mode. Facility release elevations for each gaseous release point are as indicated in Table 3-4.
8.1.1 Ground-Level Releases Relative concentration calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows:
(8.1) where:
(X/Q)G = the ground-level sector-averaged relative concentration for a given wind direction (sector) and distance (s/m 3 ).
2.032 = (2/rc)1/2 divided by the width in radians of a 22.5° sector. which is 0.3927 radians.
= the plume depletion factor for all radionuclides other than noble gases at a distance r shown in Figure 8-3. For noble gases, the depletion factor is unity.
If an undepleted relative concentration is desired. the depletion factor is unity.
Only depletion by deposition is considered since depletion by radioactive decay would be of little significance at the distances considered.
Kr = the terrain recirculation factor corresponding to a distance r. taken from Figure 8-2.
8-1 Version 24 01/10
FNP-ODCM njk = the number of hours that wind of wind speed class j is directed into the given sector during the time atmospheric stability category k existed.
N = the total hours of valid meteorological data recorded throughout the period of interest for all sectors, wind speed classes, and stability categories.
Uj = the wind speed (mid-point of wind speed class j) at ground level (m/s).
r = the distance from release point to location of interest (m).
Lzk = the vertical standard deviation of the plume concentration distribution considering the initial dispersion within the building wake, calculated as follows:
(8.2)
(jzk = the vertical standard deviation of the plume concentration distribution (m) for a given distance and stability category k as shown in Figure 8-1. The stability category is determined by the vertical temperature gradient !1T/tlz.
(OC/100 m or °F/100 ft). Plant Farley !1T/!1z values must be adjusted for tlz.
of 165 ft.
1t = 3.1416.
b = the maximum height of adjacent plant structure, which is the containment building (47 m).
8.1.2 Elevated Releases Relative dispersion calculations for elevated releases, or for the elevated portion of mixed-mode releases, shall be made as follows:
(8.3) 8-2 Version 24 01/10
FNP*ODCM where:
the elevated release sector-averaged relative concentration for a given wind direction (sector) and distance (s/m 3 ).
= the plume depletion factor for all radionuclides other than noble gases at a distance r for elevated releases, as shown in Figure 8-4, Figure 8-5, and Figure 8-6. For an elevated release, this factor is stability dependent. For noble gases, the depletion factor is unity. If an undepleted relative concentration is desired, the depletion factor is unity. Only depletion by deposition is considered since depletion by radioactive decay would be of little significance at the distances considered.
= the number of hours that wind of wind speed class j is directed into the given sector during the time atmospheric stability category k existed.
= the wind speed (mid-point of wind speed class j) at the effective release height h (m/s).
h := the effective height of the release (m), which is calculated as follows:
h hv + hpr - ht - C v (8.4)
= the height of the release point (m).
= the maximum terrain height between the release point and the point of interest (m), from Figure 2.3-26 of Reference 7.
= the additional height due to plume rise (m) which is calculated as follows and limited by hpr(max):
h pr = 1.44 d[ :: r.(
2
~r L
(8.5) 8-3 Version 24 01/10
FNP-ODCM h pr (max) the lesser of : OR (8.6) d = the inside diameter of the vent (m).
Wo = the exit velocity of the plume (m/s).
Cv = the correction for low vent exit velocity (m), which is calculated as follows:
C v =: OR (8.7)
W o for-o '?:.J.5 Uj 4 2 Fm = the momentum flux parameter (m /s ), which is calculated as follows (under the assumption that the effluent air and the ambient air have the same density):
2 Fm = (Wo )2 .(~) (8.8)
S = the stability parameter, which is calculated as follows:
S (9.S}r T
AT +9.SXJO-3)
"A z (8.9)
T = the ambient air temperature (OK).
8-4 Version 24 01/10
FNP-ODCM (AT/t1z) = the rate of increase of the ambient air temperature with increasing height above the ground (oK/m).
All other symbols are as previously defined in Section 8.1.1.
8.1.3 Mixed-Mode Releases Relative dispersion calculations for mixed-mode releases shall be made as follows:
CXIQ)M (1- E) . (XlQh + E .(XIQ)G (8.10) where:
(X/Q)M = the mixed-mode release sector-averaged relative concentration for a given wind direction (sector) and distance (s/m 3 ).
E = the fraction of hours during which releases are considered as ground-level releases, calculated as follows:
w; 1.0 for _0_ 5;, 1.0 Uj 2.58 - 1.58 . [wo J for 1.0 < wo 5;, 1.5 E= ~ ~ (8.11) fWo) Wo 0.3 0.06* -
l Uj for 1.5 < -
W; Uj 5;, 5.0 o for- o-> 5.0 Uj All other symbols are as previously defined.
8-5 Version 24 01/10
FNP-ODCM 8.2 RELATIVE DEPOSITION Plume depletion may be calculated using the appropriate form of the sector-averaged Gaussian model. Gaseous release elevations may be considered to be either at ground-level, elevated, or mixed-mode. Facility release elevations for each gaseous release points are as indicated in Table 3-4.
8.2.1 Ground-Level Releases Relative deposition calculations for ground-level releases, or for the ground-level portion of mixed-mode releases, shall be made as follows:
'IQII _ 2.55Dg Kr
~D/. /G - N 'Lnk (8.12) r k where:
(D/Q)G = the ground-level sector-averaged relative deposition for a given wind direction (sector) and distance (m-2).
2.55 = the inverse of the number of radians in a 22.5° sector [= (2 n/16r1].
Dg = the deposition rate at distance r, taken from Figure 8-7 for ground-level releases (m-1).
nk = the number of hours in which the wind is directed into the sector of interest, and during which stability category k exists.
All other symbols are as defined previously in Section 8.1.
8.2.2 Elevated Releases Relative deposition calculations for elevated releases, or for the elevated portion of mixed-mode releases, shall be made as follows:
(D/Q) = 2.55 Kr 'L(n D ) (8.13)
'E Nr k k ek where:
(D/Q)E = the elevated-plume sector-averaged relative deposition for a given wind direction (sector) and distance (m-2).
Dek = the elevated plume deposition rate at distance r, taken from Figure 8-8, Figure 8-9, or Figure 8-10, as appropriate to the plume effective release height h defined in Section 8.1.2, for stability class k (m- 1).
All other symbols are as defined previously.
8-6 Version 24 01/10
FNP-ODCM 8.2.3 Mixed-Mode Releases Relative deposition calculations for mixed-mode releases shall be made as follows:
(DIQ)M = (1- E)* (DIQ)E + E . (DIQ)G (8.14) where:
(D/Q)M:: the mixed-mode release sector-averaged relative deposition for a given wind direction (sector) and distance (m-2).
E :: the fraction of hours during which releases are considered as ground-level releases, defined in Section 8.1.3.
All other symbols are as previously defined.
8.3 ELEVATED PLUME DOSE FACTORS These factors are not required in effluent dose calculations for FNP due to the fact that all gaseous effluent releases are either ground-level or mixed-mode.
8.4 METEOROLOGICAL
SUMMARY
A summary of meteorological data for the years 1971 through 1975 is presented in Table 8-2 through Table 8-5.
8-7 Version 24 01/10
FNP-ODCM Table 8-1 Terrain Elevation Above Plant Site Grade This table intentionally left blank.
8-8 Version 24 01/10
FNP*ODCM Table 8*2 Annual Average (XJQ) for Mixed Mode Releases Distance to location, in miles Sector 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 N 2.16 E-06 9.21 E-O? 5.92 E-O? 3.83 E-O? 2.42 E-O?
NNE 2.35 E-06 1.02 E-06 6.18 E-O? 3.82 E-O? 2.34 E-O?
NE 2.23 E-06 9.61 E-O? 6.06 E-O? 3.86 E-O? 2.40 E-O?
ENE 1.12 E-06 5.03 E-O? 3.?6 E-O? 2.65 E-O? 1.?6 E-O?
E 1.20 E-06 5.21 E-O? 3.5? E-O? 2.45 E-O? 1.60 E-O?
ESE 1.55 E-06 6.43 E-O? 3.83 E-O? 2.44 E-O? 1.55 E-O?
SE 2.4? E-06 9.69 E-O? 5.52 E-O? 3.4? E-O? 2.19 E-O?
SSE 2.77 E-06 1.08 E-06 6.5? E-O? 4.34 E-O? 2.81 E-O?
S 2.50 E-06 9.3? E-O? 5.90 E-O? 4.09 E-O? 2.?4 E-O?
SSW 2.02 E-06 8.29 E-O? 6.30 E-O? 4.16 E-O? 2.66 E-O?
SW 2.05 E-06 8.34 E-O? 8.03 E-O? 5.0? E-O? 3.16 E-O?
WSW 1.89 E-06 ?41 E-O? ?33 E-O? 4.66 E-O? 2.88 E-O?
W 1.6? E-06 6.?4 E-O? 5.81 E-O? 4.12 E-O? 2.53 E-O?
WNW 1.43 E-06 5.9? E-O? 4.11 E-O? 3.13 E-O? 2.1? E-O?
NW 1.32 E-06 5.65 E-O? 3.88 E-O? 2.68 E-O? 1.77 E-O?
NNW 1.66 E-06 ?21 E-O? 4.85 E-O? 3.23 E-0?3 2.0? E-O?
Distance to location, in miles Sector 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 4.5-4.99 N 1.65 E-O? 1.24 E-O? 1.01 E-O? 9.11 E-08 8.2? E-08 NNE 1.55 E-O? 1.15E-0? 9.23 E-OB 8.28 E-08 ?4B E-08 NE 1.61 E-O? 1.19 E-O? 9.62 E-OB 8.63 E-OB ??9 E-OB ENE 1.22 E-O? 9.2B E-OB ?61 E-08 6.88 E-08 6.24 E-08 E 1.12 E-O? 8.54 E-OB ?09 E-08 6.43 E-08 5.86 E-08 ESE 1.0? E-O? 8.13 E-OB 6.?5 E-08 6.12 E-OB 5.5B E-08 SE 1.51 E-O? 1.14 E-O? 9.50 E-08 8.61 E-08 ?88 E-08 SSE 1.96 E-O? 1.50 E-O? 1.26 E-O? 1.15 E-O? 1.05 E-O?
S 1.96 E-O? 1.52 E-O? 1.29 E-O? 1.18 E-O? 1.09 E-O?
SSW 1.84 E-O? 1.39 E-O? 1.22 E-O? 1.18 E-O? 1.08 E-O?
SW 2.13 E-O? 1.60 E-O? 1.30 E-O? 1.2? E-O? 1.15 E-O?
WSW 1.92 E-O? 1.5? E-O? 1.26 E-O? 1.13 E-O? 1.02 E-O?
W 1.68 E-O? 1.69 E-O? 1.34 E-O? 1.19 E-O? 1.0B E-O?
WNW 1.?4 E-O? 1.72 E-O? 1.35 E-O? 1.21 E-O? 1.09 E-O?
NW 1.3? E-O? 1.24 E-O? 1.18 E-O? 1.06 E-O? 9.60 E-OB NNW 1.42 E-O? 1.0? E-O? 1.04 E-O? 9.36 E-08 8.50 E-08 Values are in s/m 3 , extracted from Reference 7.
8-9 Version 24 01/10
FNP-ODCM Table 8-3 Annual Average (X/Q) for Ground-Level Releases Distance to Location, in miles Sector 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 N 7.25 E-05 2.38 E-05 8.63 E-06 4.02 E-06 2.05 E-06 NNE 6.16 E-05 2.02 E-05 7.32 E-06 3.39 E-06 1.73 E-06 NE 5.86 E-05 1.94 E-05 7.04 E-06 3.24 E-06 1.65 E-06 ENE 5.27 E-05 1.74 E-05 6.32 E-06 2.92 E-06 1.49 E-06 E 6.28 E-05 2.02 E-05 7.27 E-06 3.40 E-06 1.75 E-06 ESE 6.18 E-05 1.97 E-05 7.09 E-06 3.33 E-06 1.72 E-06 SE 9.48 E-05 3.01 E-05 1.07 E-05 5.06 E-06 2.63 E-06 SSE 1.44 E-04 4.55 E-05 1.61 E-05 7.65 E-06 3.99 E-06 S 1.55 E-04 4.87 E-05 1.72 E-05 8.20 E-06 4.28 E-06 SSW 9.78 E-05 3.12 E-05 1.11 E-05 5.23 E-06 2.71 E-06 SW 7.40 E-05 2.40 E-05 8.74 E-06 4.05 E-06 2.07 E-06 WSW 6.01 E-05 1.97 E-05 7.18 E-06 3.31 E-06 1.68 E-06 W 5.76 E-05 1.88 E-05 6.79 E-06 3.14 E-06 1.60 E-06 WNW 5.55 E-05 1.82 E-05 6.55 E-06 3.03 E-06 1.55 E-06 NW 5.67 E-05 1.86 E-05 6.76 E-06 3.14 E-06 1.60 E-06 NNW 6.60 E-05 2.16 E-05 7.85 E-06 3.65 E-06 1.87 E-06
~
Distance to Location, in miles Sector 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 N 1.19 E-06 8.24 E-07 6.09 E-07 5.35 E-07 4.71 E-07 NNE 1.00 E-06 6.94 E-07 5.13E-07 4.50 E-07 3.96 E-07 NE 9.47 E-07 6.54 E-07 4.82 E-07 4.23 E-07 3.71 E-07 ENE 8.56 E-07 5.92 E-07 4.37 E-07 3.82 E-07 3.37 E-07 E 1.02 E-06 7.08 E-07 5.24 E-07 4.61 E-07 4.06 E-07 ESE 1.02 E-06 6.99 E-07 5.18E-07 4.56 E-07 4.02 E-07 SE 1.54 E-06 1.07 E-06 7.99 E-07 7.04 E-O? 6.20 E-O?
SSE 2.34 E-06 1.64 E-06 1.22 E-06 1.08 E-06 9.49 E-07 S 2.51 E-06 1.76 E-06 1.31 E-06 1.16 E-06 1.02 E-06 SSW 1.58 E-06 1.10 E-06 8.17 E-07 ?19 E-O? 6.33 E-07 SW 1.20 E-06 8.30 E-07 6.12 E-07 5.38 E-O? 4.73 E-07 WSW 9.65 E-O? 6.67 E-07 4.91 E-07 4.31 E-O? 3.?9 E-O?
W 9.20 E-07 6.37 E-07 4.71 E-O? 4.13 E-O? 3.63 E-07 WNW 8.92 E-07 6.18 E-07 4.56 E-07 4.01 E-O? 3.52 E-07 NW 9.25 E-07 6.41 E-07 4.73 E-07 4.16E-07 3.65 E-07 NNW 1.10 E-06 7.50 E-07 5.54 E-07 4.8? E-07 4.28 E-07 Values are in 51m 3 , extracted from Reference 7.
8-10 Version 24 01/10
FNP*ODCM Table 8*4 Annual Average (D/Q) for Mixed Mode Releases Distance to Location, in miles
<::ector 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49~1 N 3.B2 E-OB 1.7B E-OB 7.53 E-09 3.39 E-09 1.62 E-09 NNE 4.57 E-OB 2.0B E-OB B.69 E-09 3.BB E-09 1.B5 E-09 NE 4.7B E-OB 2.20 E-OB 9.0B E-09 4.03 E-09 1.92 E-09 ENE 2.67 E-OB 1.32 E-OB 5.63 E-09 2.54 E-09 1.22 E-09 E 2.B7 E-OB 1.40 E-OB 5.77 E-09 2.55 E-09 1.22 E-09 ESE 3.29 E-OB 1.53 E-OB 6.17 E-09 2.70 E-09 1.2B E-09 SE 5.30 E-OB 2.37 E-OB 9.31 E-09 4.01 E-09 1.90 E-09 SSE 5.07 E-OB 2.35 E-OB 9.53 E-09 4.19E-09 1.99 E-09 S 4.B6 E-OB 2.29 E-OB 9.16 E-09 4.00 E-09 1.90 E-09 SSW 4.29 E-OB 2.10 E-OB 9.09 E-09 3.97 E-09 1.BB E-09 SW 4.70 E-OB 2.2B E-OB 1.05 E-OB 4.39 E-09 2.04 E-09 WSW 4.46 E-OB 2.17 E-OB 9.BB E-09 4.12 E-09 1.92 E-09 W 3.96 E-OB 1.94 E-OB B.39 E-09 3.63 E-09 1.70 E-09 WNW 3.22 E-OB 1.56 E-OB 6.35 E-09 2.B5 E-09 1.37 E-09 NW 2.B3 E-08 1.35 E-OB 5.55 E-09 2.46 E-09 1.18 E-09 NNW 3.24 E-08 1.55 E-08 6.59 E-09 2.97 E-09 1.42 E-09 Distance to Location in miles Sector
. -3.49 4.0-4.49 N 8.71 E-10 5.64 E-10 3.10 E-10 3.37 E-10 2.91 E-10 NNE 9.91 E-10 6.43 E-10 4.44 E-10 3.82 E-10 3.30 E-10 NE 1.03 E-09 6.65 E-10 4.62 E-10 3.9B E-10 3.43 E-1 0 ENE 6.57 E-10 4.22 E-10 2.96 E-10 2.55 E-10 2.20 E-10 E 6.57 E-10 4.20 E-10 2.96 E-10 2.55 E-10 2.20 E-10 ESE 6.BB E-10 4.40 E-10 3.09 E-10 2.66 E-10 2.29 E-10 SE 1.01 E-09 6.48 E-10 4.55 E-10 3.90 E-10 3.36 E-10 SSE 1.07 E-09 6.B5 E-10 4.79 E-10 4.12E-10 3.55 E-10 S 1.02 E-09 6.49 E-10 4.59 E-10 3.94 E-10 3.40 E-10 SSW 1.00 E-09 6.41 E-10 4.50 E-10 3.B6 E-10 3.32 E-10 SW 1.0B E-09 6.90 E-10 4.B1 E-10 4.12 E-10 3.53 E-10 WSW 1.02 E-09 6.51 E-10 4.53E-10 3.B7 E-10 3.32 E-10 W 9.00 E-10 5.92 E-10 4.13E-10 3.54 E-10 3.04 E-10 WNW 7.33 E-10 4.95 E-10 3.52 E-10 3.05 E-10 2.65 E-10 NW 6.37 E-10 4.11 E-10 2.91 E-10 2.50 E-10 2.14 E-10 NNW 7.66 E-10 4.95 E-10 3.45 E-10 2.97 E-10 2.56 E-10 Values are in m*2 , extracted from Reference 7.
8-11 Version 24 01/10
FNP-ODCM Table 8-5 Annual Average (0/0) for Ground-Level Releases Sector Distance to Location, in miles II 0.25-0.5 0.5-0.99 1.0-1.49 1.5-1.99 2.0-2.49 II N 2.50 E-07 7.B4 E-OB 2.53 E-08 9.61 E-09 4.2B E-09 NNE 2.4B E-07 7.77 E-OB 2.51 E-OB 9.53 E-09 4.24 E-09 NE 2.49 E-07 7.BO E-OB 2.52 E-OB 9.57 E-09 4.26 E-09 ENE 1.69 E-07 5.29 E-OB 1.71 E-OB 6.4B E-09 2.BB E-09 E 1.69 E-07 5.28 E-OB 1.71 E-OB 6.4B E-09 2.BB E-09 ESE 1.BO E-07 5.54 E-08 1.79E-OB 6.BO E-09 3.02 E-09 SE 2.75 E-07 B.63 E-OB 2.79 E-OB 1.06 E-OB 4.71 E-09 SSE 3.66 E-07 1.15 E-07 3.71 E-OB 1.41 E-OB 6.25 E-09 S 3.70 E-07 1.16E-07 3.75 E-OB 1.42 E-OB 6.33 E-09 SSW 2.75 E-07 B.62 E-OB 2.79 E-OB 1.06 E-OB 4.70 E-09 SW 2.60 E-07 B.15 E-OB 2.64 E-OB 1.00 E-OB 4.45 E-09 WSW 2.31 E-07 7.24 E-OB 2.34 E-OB B.BB E-09 3.95 E-09 W 2.11 E-07 6.61 E-OB 2.14 E-OB B.11 E-09 3.61 E-09 WNW 1.83 E-07 5.73 E-OB 1.85 E-OB 7.02 E-09 3.12 E-09 NW 1.74 E-07 5.45 E-OB 1.76 E-08 6.6B E-09 2.97 E-09 NNW 2.13 E-07 6.67 E-OB 2.16 E-08 B.19 E-09 3.64 E-09 C'
Distance to Location, in miles 2.5-2.99 3.0-3.49 3.5-3.99 4.0-4.49 4.5-4.99~1 N 2.22 E-09 1.45 E-09 9.79 E-l0 8.27 E-l0 6.99 E-10 NNE 2.20 E-09 1.43 E-09 9.71 E-10 8.20 E-10 6.93 E-10 NE 2.21 E-09 1.44 E-09 9.75 E-10 B.23 E-10 6.96 E-10 ENE 1.50 E-09 9.76 E-10 6.60 E-10 5.5B E-10 4.72 E-10 E 1.50 E-09 9.75 E-10 6.60 E-10 5.57 E-10 4.71 E-10 ESE 1.57 E-09 1.02 E-09 6.72 E-10 5.B5 E-10 4.94 E-10 SE 2.44 E-09 1.59 E-09 1.08 E-10 9.11 E-10 7.70 E-10 SSE 3.25 E-09 2.12 E-09 1.43E-10 1.21 E-l0 1.02 E-l0 S 3.29 E-09 2.14 E-09 1.45 E-10 1.22 E-09 1.04 E-l0 SSW 2.44 E-09 1.59 E-09 1.08 E-l0 9.10 E-10 7.69 E-10 SW 2.31 E-09 1.51 E-09 1.02 E-10 B.60 E-10 7.27 E-10 WSW 2.05 E-09 1.34 E-09 9.04 E-10 7.64 E-10 6.46 E-10 W 1.87 E-09 1.22 E-09 B.25E-l0 6.97 E-l0 5.90 E-l0 WNW 1.62 E-09 1.06 E-09 7.15 E-10 6.04 E-10 5.11 E-10 NW 1.54 E-09 1.01 E-09 6.BO E-10 5.75 E-l0 4.B6 E-10 NNW 1.B9 E-09 1.23 E-09 8.34 E-10 7.04 E-10 5.95 E-10 Values are in m*2 , extracted from Reference 7.
8-12 Version 24 01/10
FNP*ODCM 1000 I I II I /"
/
'" "" ./
v I / )"" L 100 AI/
V I /V V r-/V
..-I- -
/ V /'
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I'
/ /
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./
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1 0.1 1.0 10 100 PLUME TRAVEL DISTANCE (KILOMETERSI Range of Vertical Temperature Range of Vertical Temperature i Category Gradient (OC/100 m) Gradient (OF/100 ft)
A AT/il.< -1.9 AT/AZ < -1.0 B -1.9::;AT/il.< -1.7 -1.0::; AT/il. < -0.9 C -1.7 < AT/AZ < -1.5 -0.9::; AT/AZ < -0.8 D -1.5::; AT/AZ < -0.5 -0.8::; AT/AZ < -0.3 E -0.5::; AT/il. < 1.5 -0.3::; AT/AZ < 0.8 F 1.5::; AT/il. < 4.0 0.8::; AT/il. < 2.2 G 4.0 ::;AT/AZ 2.2 ::;AT/AZ This graph is reproduced from Reference 5 (Figure 1).
Figure 8*1 Vertical Standard Deviation of Material in a Plume (O"z) 8-13 Version 24 01/10
FNP-ODCM
[use former Figure 3-3 or comparable]
to
, I j I I I I T 1 I 1 1I I I I I I I I I i I 111111 " I I I I IIIII '"t'-.
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Figure 8-2 Terrain Recirculation Factor (Kr) 8-14 Version 24 01/10
FNP-ODCM i
I I I 1.0 !
0.9 I
~ ..
~
I i
...::\! O.B I 1 i" r-. .... I
...:= ,i II.
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i !
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0.1 I I I I I I
i I
0.1 1.~ 10.0 tOO.O 200.0 PLUME TRAVEL OISTANCE (KILOMETERS!
This graph is reproduced from Reference 5 (Figure 2).
Figure 8*3 Plume Depletion Effect for Ground-Level Releases 8-15 Version 24 01/10
FNP*ODCM I
1.0 K.A8lE w
0.1 o.a F=::::::--- l--.. NEUTRALlDI
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0.1 1.0 10.0 100J1' aoo.o PLUME TRAVEL DISTANCE IKILOMnERII This graph is reproduced from Reference 5 (Figure 3).
Figure 8-4 Plume Depletion Effect for 30-Meter Releases 8-16 Version 24 01/10
FNP-ODCM 1.0 I STABLE
-:::::: :::.: :::: ~ ....
~Q) 0.9
.... O.B
~
........ "" ~EUTRAL'
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U 0.3 un ~
0.2
\
i 0.1 0.1 I
1.0 10.0 1
100.0 200.0 PLUME TRAVEL DISTANCE (KILOMETERS)
This graph is reproduced from Reference 5 (Figure 4).
Figure 8-5 Plume Depletion Effect for 60-Meter Releases 8-17 Version 24 01/10
FNP-ODCM 1.0 r-- r-- r- r- ..... ~
NEUTRAL(DI
.... ~
~ .......... ~
UNSTABLE 1A.B,c1 "- i'r-..
~
STABLE (E.F.OI NO DEPLETION (FRACTION REMAINING -1.0) \
1'1 "
\.
0.3 \
Q.2
- \
0.1 0.1 1.0 10.0 100.0 200.0 PLUME TRAVEL DISTANCE (KILOMETERS!
This graph is reproduced from Reference 5 (Figure 5).
Figure 8-6 Plume Depletion Effect for 1DO-Meter Releases 8-18 Version 24 01/10
FNP*ODCM I I
~
w 10-4 I
w
i ""'
a:
w e:. ""
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en
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10-6 ~
10-7 0.1 1.0 10.0 100.0 200.0 PLUME TRAVEL DISTANCE (KILOMETERS)
This graph is reproduced for Reference 5 (Figure 6).
Figure 8-7 Relative Deposition for Ground-Level Releases 8-19 Version 24 01/10
FNP*ODCM 10-3 U~ABLE (A.B,c1 a:
t 10-4
i a: ~ "" ~
..... II'
'""0" NEUTRAL t- j o(
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10-6 U

i 0
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~
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..... I .......
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..... I "- ..,.
~
a:
10-6 / I ISTABLE (E.F.GI I
I
~
10-7 0.1 1.0 10.0 100..0 200..0 PLUME TRAVEL DISTANCE (KILOMETERS' This graph is reproduced from Reference 5 (Figure 7).
Figure 8*8 Relative Deposition for 30-Meter Releases 8-20 Version 24 01/10
FNP*ODCM
... UNSTABLE tA. B.C)
V '
/
V "' b..
~
~ !
lcr6 / NEUTRAL (D) ~=:
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I I
II 10-8 l 0.1 1~ 10~ 100.0 200~
PLUME TRAVEL DISTANCE (KILOMETERS)
This graph is reproduced from Reference 5 (Figure 8).
Figure 8*9 Relative Deposition for 50-Meter Releases 8-21 Version 24 01/10
FNP*ODCM 10-4 I~
UNSTABLE (A.B,C)
I V
i'- ,
2
,0-5 .'
I
! NEUTRAL CO) ill I
w """'"......
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w II I' e:.
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~
t o(
a:
2 0
j:
1ct 1, the maximum permissible effluent discharge flowrate for this release pathway, fmp (in gpm), is calculated as follows:
Fdp Imp = (RDF -1) (2.7)
For the case RDF ~ 1, equation (2.7) is not valid. However, as discussed above, when RDF ~ 1, the release may be made at full discharge pump capacity; the radioactivity monitor setpoint must still be calculated in accordance with Step 5 below.
NOTE 1: Discharge flowrates are actually limited by the discharge pump capacity. When the calculated maximum permissible release flowrate exceeds the pump capacity, the release may be made at full capacity. Discharge flowrates less than the pump capacity must be achieved by throttling if this is available; if throttling is not available, the release may not be made as planned.
NOTE 2: If, at the time of the planned release, there is detectable radioactivity due to plant operations in the dilution stream, the diluting capacity of the dilution stream is diminished. (In addition, sampling and analysis of the other radioactive effluents affecting the dilution stream must be sufficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.) Under these conditions, equation (2.7) must be modified to account for the radioactivity present in the dilution stream prior to the introduction of the planned release:
2-22 VER27
VEGPODCM t,np (2.8) where:
= the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution stream.
= the effluent discharge flowrate of release pathway r.
If the entire dilution stream contains detectable activity due to plant operations, whether or not its source is identified, fr = Fd , and Cir is the concentration in the total dilution system. This note does not apply: a) if the RDF of the planned release is S 1; or b) if the release contributing radioactivity to the dilution stream has been accounted for by the assignment of an allocation factor.
Step 5: Determine the maximum radioactivity monitor setpoint concentration.
Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the limits of Section 2.1.2 will not be exceeded.
Because the radioactivity monitor responds primarily to gamma radiation, the monitor setpoint cp for release pathway p (in IlCi/mL) is based on the concentration of gamma emitters in the waste stream, as follows:
(2.9) where:
Ap = an adjustment factor which will allow the setpoint to be established in a practical manner to prevent spurious alarms while allowing a margin between measured concentrations and the limits of Section 2.1.2.
Step 5.a. If the concentration of gamma emitters in the effluent to be released is sufficient that the high alarm setpoint can be established at a level that will prevent spurious alarms, Ap should be calculated as follows:
A =_l-xADF p RDF (2.10) 1 (F + tap) dP
= - - x -'----'-
RDF iap where:
ADF = the assured dilution factor.
= the anticipated actual discharge flowrate for the planned release (in gpm), a value less than fmp
The release must then be controlled so that the actual efl'luent discharge flowrate does not exceed fap at anytime.
2-23 VER27
VEGPODCM Step 5.b.Alternatively, Ap may be calculated as follows:
ADF - RDFny RDFy (2.11 )
Step 5.c. Evaluate the computed value of Ap as follows:
If Ap ;:::: 1, calculate the monitor setpoint, cpo However, if cp is within about 10 percent of Cg , it may be impractical to use this value of cpo This situation indicates that measured concentrations are approaching values which would cause limits of Section 2.1.2 to be exceeded.
Therefore, steps should be taken to reduce potential concentrations at the point of discharge; these steps may include decreasing the planned effluent discharge flowrate, increasing the dilution stream flowrate, postponing simultaneous releases, and/or decreasing the effluent concentrations by further processing the liquid planned for release. Alternatively, allocation factors for the active liquid release pathways may be reassigned. When one or more of these actions has been taken, repeat Steps 1-5 to calculate a new radioactivity monitor setpoint.
If Ap < 1, the release may not be made as planned. Consider the alternatives discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.
2.3.2.3 Use of the Calculated Setpoint The setpoint calculated above is in the units J!CilmL. The monitor actually measures a count rate, subtracts a predetermined background count rate, and multiplies by a calibration factor to convert from count rate to IlCi/mL.
Initial calibration of the monitors by the manufacturer and Georgia Power Company utilized NIST traceable liquid solutions with gamma ray emissions over the range 0.08 to 1.33 MeV, in the exact geometry of each production monitor. The calibration factor is a function of the radionuclide mix in the liquid to be released, and will be calculated for the monitor based on the results of the pre release sample results from the laboratory gamma-ray spectrometer system. The mix-dependent calibration factor will be used as the gain factor in the PERMS monitor, or used to modify the calculated base monitor setpoint so that the default calibration factor in the PERMS monitor can be left unchanged.
Notwithstanding the initial calibration, monitor calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to stream concentrations measured by liquid sample analysis. In all cases, monitor background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value.
2-24 VER27
VEGP ODCM 2.3.3 Setpoints for Monitors on Normally Low-Radioactivity Streams Radioactivity in these streams (listed in Table 2-4 above) is expected to be at very low levels, generally below detection limits. Accordingly, the purpose of these monitors is to alarm upon the occurrence of significant radioactivity in these streams, and to terminate or divert the release where this is possible.
2.3.3.1 Normal Conditions When radioactivity in one of these streams is at its normal low level, its radioactivity monitor setpoint should be established as close to background as practical to prevent spurious alarms, and yet alarm should an inadvertent release occur.
2.3.3.2 Conditions Requiring an Elevated Setpoint Under the following conditions, radionuclide concentrations must be determined and an elevated radioactivity monitor setpoint determined for these pathways:
For streams that can be diverted or isolated, a new monitor setpoint must be established when it is desired to discharge the stream directly to the dilution water even though the radioactivity in the stream exceeds the level which would normally be diverted or isolated.
For streams that cannot be diverted or isolated, a new monitor setpoint must be established whenever: the radioactivity in the stream becomes detectable above the background levels of the applicable laboratory analyses; or the associated radioactivity monitor detects activity in the stream at levels above the established alarm setpoint.
When an elevated monitor setpoint is required for any of these effluent streams, it should be determined in the same manner as described in Section 2.3.2. However, special consideration must be given to Step 3. An allocation factor must be assigned to the normally low-radioactivity release pathway under consideration. and allocation factors for other release pathways discharging simultaneously must be adjusted downward (if necessary) to ensure that the sum of the allocation factors does not exceed 1. Sampling and analysis of the normally low-radioactivity streams must be sufficient to ensure that the liquid effluent dose limits specified in the controls of Section 2.1.3 are not exceeded.
2-25 VER27
VEGPODCM 2.4 LIQUID EFFLUENT DOSE CALCULATIONS The following sub-sections present the methods required for liquid effluent dose calculations, in deepening levels of detail. Applicable site-specific pathways and parameter values for the calculation of D~, Ai~' and CF1v are summarized in Table 2-5.
2.4.1 Calculation of Dose The dose limits for a MEMBER OF THE PUBLIC specified in Section 2.1.3 are on a per-unit basis. Therefore, the doses calculated in accordance with this section must be determined and recorded on a per-unit basis, including apportionment of releases shared between the two units.
For the purpose of implementing Section 2.1.3, the dose to the maximum exposed individual due to radionuclides identified in liquid effluents released from each unit to UNRESTRICTED AREAS will be calculated as follows (equation from Reference 1, page 15):
(2.12) where:
D~ = the cumulative dose commitment to the total body or to any organ 't, in mrem, due to radioactivity in liquid effluents released during the total of the m time periods Atl
Ait = the site-related adult ingestion dose commitment factor, for the total body or for any organ 1:, due to identified radionuclide i, in (mrem*mL)/(h*IlCi). Methods for the calculation of Ai't are presented below in Section 2.4.2. The values of At to be used in dose calculations for releases from the plant site are listed in Table 2-8.
Atl = the length of time period I, over which Cil and FI are averaged for liquid releases, in h.
Cil = the average concentration of radionuclide i in undiluted liquid effluent during time period I, in J.LCi/mL. Only radionuclides identified and detected above background in their respective samples should be included in the calculation.
F, = the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA:
F==~ (2.13)
I Ft xZ where:
ft = the average undiluted liquid waste flowrate actually observed during the period of radioactivity release, in gpm.
Ft = the average dilution stream flowrate actually observed during the period of radioactivity release, in gpm. If simultaneous releases from both units occur, 2-26 VER27
VEGPODCM the dilution stream flowrate Ft must be allocated between them. In such cases, F, is unit~specific.
z = the applicable dilution factor for the receiving water body, in the near field of the discharge structure, during the period of radioactivity release, from Table 2-5.
NOTE: In equation (2.13), the product (Ft x Z) is limited to 1000 cfs
(= 448,000 gpm) or less. (Reference 1, Section 4.3.)
2.4.2 Calculation of Ait The site~related adult ingestion dose commitment factor, Ait, is calculated as follOWS (equation adapted from Reference 1, page 16, by addition of the irrigated garden vegetation pathway):
(2.14) where:
1.14x105 = a units conversion factor, determined by:
Uw = the adult drinking water consumption rate applicable to the plant site (Uy).
Dw = the dilution factor from the near field of the discharge structure for the plant site to the potable water intake location.
Ai = the decay constant for radionuclide i (h-1). Values of Ai used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 20.
fw = the transit time from release to receptor for potable water consumption (h).
Uf = the adult rate of fish consumption applicable to the plant site (kg/y).
SFi = the bioaccumulation factor for radionuclide i applicable to freshwater fish in the receiving water body for the plant site, in (pCi/kg)/(pCi/L) = (Ukg). For specific values applicable to the plant site, see Table 2~6.
tf = the transit time from release to receptor for fish consumption (h).
Uv = the adult consumption rate for irrigated garden vegetation applicable to the plant site (kg/y).
CFjv = the concentration factor for radionuclide i in irrigated garden vegetation, as applicable to the vicinity of the plant site, in (pCi/kg)/(pCi/L). Methods for calculation of CFiv are presented below in Section 2.4.3.
2-27 VER27
VEGPODCM DFit = the dose conversion factor for radionuclide i for adults, in organ t (mrem/pCi). For specific values, see Table 2-7.
2.4.3 Calculation of Cfiv The concentration factor for radionuclide i in irrigated garden vegetation, CFiv in (Ukg), is calculated as follows:
For radionuclides other than tritium (equation adapted from Reference 3, equations A-8 and A-9):
(
CF =M.J r1-e E"+
-A J) f I BIv (1 -e -Alb)]
' -V. (2.15)
Iv [ Y1 v EI Pl i
e
For tritium (equation adapted from Reference 3, equations A-9 and A-10):
(2.16) where:
M = the additional river dilution factor from the near field of the discharge structure for the plant site to the point of irrigation water usage.
I = the average irrigation rate during the growing season (L)/(m2 *h).
r = the fraction of irrigation-deposited activity retained on the edible portions of leafy garden vegetation.
Yv = the areal density (agricultural productivity) of leafy garden vegetation (kg/m2) f, = the fraction of the year that garden vegetation is irrigated.
Biv = the crop to soil concentration factor applicable to radionuclide i (pCi/kg garden vegetation )/(pCi/kg soil).
P = the effective surface density of soil (kglm1.
Ai = the decay constant for radionuclide i (h- 1). Values of Ai used in effluent calculations should be based on decay data from a recognized and current source, such as Reference 20.
'A.w = the rate constant for removal of activity from plant leaves by weathering (h-1).
AEi = the effective removal rate for activity deposited on crop leaves (h-1) calculated as:
As =Ai + 'A.w.
te = the period of leafy garden vegetation exposure during the growing season (h).
tb = the period of long-term buildup of activity in soil (h).
2-28 VER27
VEGPODCM th = the time between harvest of garden vegetation and human consumption (h).
Lv = the water content of leafy garden vegetation edible parts (Ukg).
2-29 VER27
VEGPODCM Table 2-5. Parameters for Calculation of Doses Due to Liquid Effluent Releases Dose Calculation Receptor Locations:
Fish: Vicinity of plant discharge Drinking Water: 112 miles downstream, at Beaufort, SC (Reference 12)
Irrigated Garden Vegetation: None (Reference 12)
Numerical Parameters:
Parameter Value Reference Z 10, for May through December Ref. 11 20, for January through April Uw 730Uy Ref 3 Dw 8 Ref. 7 tw 48h Ref. 3, Sec. A.2; Ref. 8 Uf 21 kg/y Ref. 3, Table E-5 tf 24 h Ref. 3, Sec. A.2 Uy o kg/y Ref. 12 M 1.0+
I No value r 0.25 Ref. 3, Table E-15.
Yy 2.0 kg/m2 Ref. 3, Table E-15 f/ 1.0 +
P 240 kg/m2 Ref. 3, Table E-15 Aw 0.0021 h-1 (Le., half-life of 14 d) Ref. 3, Table E-15 te 1440 h (= 60 d) Ref. 3, Table E-15 it, 1.31 x 105 h (= 15 y) Ref. 3, Table E-15 tt, 24 h Ref. 3, Table E-15 Lv 0.92 Ukg Based on Ref. 21, Table 5.16 (for lettuce, cabbage, etc.)
Because there is no irrigated garden vegetation pathway downstream of the plant site, the consumption of irrigated garden vegetation is set to zero, and the other pathway parameters are defaults.
+ - There is no established default value for this parameter. The most conservative physically realistic value is 1.0.
2-30 VER27
VEGPODCM Table 2-6. Element Transfer Factors Freshwater Fish Element BFi H 9.0 E-01 C 4.6 E+03 Na 1.0 E+02 P 3.0 E+03 Cr 2.0 E+02 Mn 4.0 E+02 Fe 1.0 E+02 Co 5.0 E+01 Ni 1.0 E+02 Cu 5.0 E+01 Zn 2.0 E+03 Br 4.2 E+02 Rb 2.0 E+03 Sr 3.0 E+01 y 2.5 E+01 Zr 3.3 E+OO Nb 5.5 E+02 Mo 1.0 E+01 Tc 1.5 E+01 Ru 1.0 E+01 Rh 1.0 E+01 Ag 2.3 E+OO Sb 2.0 E+02 Te 4.0 E+02 I 1.5 E+01 Cs 2.0 E+03 8a 4.0 E+OO La 2.5 E+01 Ce 1.0 E+OO Pr 2.5 E+01 Nd 2.5 E+01 W 1.2 E+03 Np 1.0 E+01
Bioaccumulation Factors for freshwater fish, in (pCi/kg)/(pCi/L). They are obtained from Reference 3 (Table A-1), except as follows: Reference 9 for P; Reference 2 (Table A-B) for Ag; and Reference 10 for Nb and Sb.
2-31 VER27
VEGPODCM Table 2-7. Adult Ingestion Dose Factors I~ H-3 B~"or T.Body Thyroid Kidney Lung GI-LLI No Data 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 1.05E-07 C-14 2.B4E-06 5.6BE-07 5.6BE-07 5.6BE-07 5.6BE-07 5.6BE-07 5.6BE-07 Na-24 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 1.70E-06 P-32 1.93E-04 1.20E-05 7A6E-06 No Data No Data No Data 2.17E-05 Cr-51 No Data No Data 2.66E-09 1.59E-09 5.B6E-l0 3.53E-09 6.69E-07 Mn-54 No Data 4.57E-06 B.72E-07 No Data 1.36E-06 No Data 1.40E-05 Mn-56 No Data 1.15E-07 2.04E-OB No Data lA6E-07 No Data 3.67E-06 Fe-55 2.75E-06 1.90E-06 4.43E-07 No Data No Data 1.06E-06 1.09E-06 Fe-59 4.34E-06 1.02E-05 3.91 E-06 No Data No Data 2.B5E-06 . 3.40E-05 Co-5B No Data 7.45E-07 1.67E-06 No Data No Data No Data 1.51 E-05 Co-60 No Data 2.14E-06 4.72E-06 No Data No Data No Data 4.02E-05 Ni-63 1.30E-04 9.01 E-06 4.36E-06 No Data No Data No Data 1.BBE-06 Ni-65 5.2BE-07 6.B6E-OB 3.13E-OB No Data No Data No Data 1.74E-06 Cu-64 No Data B.33E-OB 3.91E-OB No Data 2.10E-07 No Data 7. 1OE-06 Zn-65 4.B4E-06 1.54E-05 6.96E-06 No Data 1.03E-05 No Data 9.70E-06 Zn-69 1.03E-OB 1.97E-OB 1.37E-09 No Data 1.2BE-OB No Data 2.96E-09 Br-B3 No Data No Data 4.02E-OB No Data No Data No Data 5.79E-OB Br-B4 No Data No Data 5.21 E-OB No Data No Data No Data 4.09E-13 Br-B5 No Data No Data 2.14E-09 No Data No Data No Data No Data Rb-B6 No Data 2.11 E-05 9.B3E-06 No Data No Data No Data 4.16E-06 Rb-BB No Data 6.05E-OB 3.21E-OB No Data No Data No Data B.36E-19 Rb-S9 No Data 4.01E-OS 2.S2E-OS No Data No Data No Data 2.33E-21 Sr-B9 3.0BE-04 No Data B.B4E-06 No Data No Data No Data 4.94E-05 Sr-90 7.5BE-03 No Data 1.S6E-03 No Data No Data No Data 2.19E-04 Sr-91 5.67E-06 No Data 2.29E-07 No Data No Data No Data 2.70E-05 All values are in (mrem/pCi ingested). They are obtained from Reference 3 (Table E-11).
except as follows: Reference 2 (Table A-3) for Rh-105, Sb-124, and Sb-125.
2-32 VER27
VEGPODCM Table 2-7 (contd). Adult Ingestion Dose Factors I~ Sr-92 Bone 2.15E-06 Liver T.Body Thyroid Kidney I Lung GI-LLI No Data 9.30E-08 No Data No Data No Data 4.26E-05 Y-90 9.62E-09 No Data 2.58E-10 No Data No Data No Data 1.02E-04 Y-91m 9.09E-11 No Data 3.52E-12 No Data No Data No Data 2.67E-10 Y-91 1.41E-07 No Data 3.77E-09 No Data No Data No Data 7.76E-05 Y-92 8,45E-10 No Data 2,47E-11 No Data No Data No Data 1,48E-05 Y-93 2.68E-09 No Data 7,40E-11 No Data No Data No Data 8.50E-05 Zr-95 3.04E-08 9.75E-09 6.60E-09 No Data 1.53E-08 No Data 3.09E-05 Zr-97 1.68E-09 3.39E-10 1.55E-10 No Data 5.12E-10 No Data 1.05E-04 Nb-95 6.22E-09 3,46E-09 1.86E-09 No Data 3,42E-09 No Data 2.10E-05 Mo-99 No Data 4.31 E-06 8.20E-07 No Data 9.76E-06 No Data 9.99E-06 Tc-99m 2,47E-10 6.98E-10 8.89E-09 No Data 1.06E-08 3,42E-10 4.13E-07 Tc-101 2.54E-10 3.66E-10 3.59E-09 No Data 6.59E-09 1.87E-10 1.10E-21 Ru-103 1.85E-07 No Data 7.97E-08 No Data 7.06E-07 No Data 2.16E-05 Ru-105 1.54E-08 No Data 6.08E-09 No Data 1.99E-07 No Data 9,42E-06 Ru-106 2.75E-06 No Data 3,48E-07 No Data 5.31E-06 No Data 1.78E-04 Rh-105 1.22E-07 8.86E-08 5.83E-08 No Data 3.76E-07 No Data 1,41E-05 Ag-110m 1.60E-07 1,48E-07 8.79E-08 No Data 2.91E-07 No Data 6.04E-05 Sb-124 2.81E-06 5.30E-08 1.11 E-06 6.79E-09 No Data 2.18E-06 7.95E-05 Sb-125 2.23E-06 2,40E-08 4,48E-07 1.98E-09 No Data 2.33E-04 1.97E-05 Te-125m 2.68E-06 9.71E-07 3.59E-07 8.06E-07 1.09E-05 No Data 1.07E-05 Te-127m 6.77E-06 2,42E-06 8.25E-07 1.73E-06 2.75E-05 No Data 2.27E-05 Te-127 1.10E-07 3.95E-08 2.38E-08 8.15E-08 4,48E-07 No Data 8.68E-06 Te-129m 1.15E-05 4.29E-06 1.82E-06 3.95E-06 4.80E-05 No Data 5.79E-05 Te-129 3.14E-08 1.18E-08 7.65E-09 2.41E-08 1.32E-07 No Data 2.37E-OS Te-131 m 1.73E-06 S,46E-07 7.05E-07 1.34E-06 S.57E-06 No Data 8,40E-05 Te-131 1.97E-OB B.23E-09 6.22E-09 1.62E-OS S.63E-OS No Data 2.79E-09 2-33 VER27
VEGPODCM Table 2-7 (contd). Adult Ingestion Dose Factors Bone Liver T.Body Thyroid Kidney Lung GI-LLI Te-132 2.S2E-06 1.63E-06 1.S3E-06 1.BOE-06 1.S7E-OS No Data 7.71 E-OS 1-130 7.S6E-07 2.23E-06 B.BOE-07 1.B9E-04 3.4BE-06 No Data 1.92E-06
'-131 4.16E-06 S.9SE-06 3.41E-06 1.9SE-03 1.02E-OS No Data 1.S7E-06 1-132 2.03E-07 S.43E-07 1.90E-07 1.90E-OS B.6SE-07 No Data 1.02E-07 1-133 1.42E-06 2.47E-06 7.S3E-07 3.63E-04 4.31E-06 No Data 2.22E-06 1-134 1.06E-07 2.BBE-07 1.03E-07 4.99E-06 4.SBE-07 No Data 2.S1 E-10 1-13S 4.43E-07 1.16E-06 4.2BE-07 7.6SE-OS 1.B6E-06 No Data 1.31E-06 Cs-134 6.22E-OS 1.4BE-04 1.21E-04 No Data 4.79E-OS 1.S9E-OS 2.S9E-06 Cs-136 6.S1 E-06 2.S7E-OS 1.BSE-OS No Data 1.43E-OS 1.96E-06 2.92E-06 Cs-137 7.97E-OS 1.09E-04 7.14E-OS No Data 3.70E-OS 1.23E-OS 2.11E-06 Cs-13B S.S2E-OB 1.09E-07 S.40E-OB No Data B.01 E-OB 7.91 E-09 4.6SE-13 Ba-139 9.70E-OB 6.91 E-11 2.B4E-09 No Data 6.46E-11 3.92E-11 1.72E-07 Ba-140 2.03E-OS 2.SSE-OB 1.33E-06 No Data B.67E-09 1.46E-OB 4.1BE-OS Ba-141 4.71E-OB 3.S6E-11 1.S9E-09 No Data 3.31E-11 2.02E-11 2.22E-17 Ba-142 2.13E-OB 2.19E-11 1.34E-09 No Data 1.85E-11 1.24E-11 3.00E-26 La-140 2.S0E-09 1.26E-09 3.33E-10 No Data No Data No Data 9.2SE-OS La-142 1.2BE-10 S.B2E-11 1.4SE-11 No Data No Data No Data 4.2SE-07 Ce-141 9.36E-09 6.33E-09 7.1BE-10 No Data 2.94E-09 No Data 2.42E-OS Ce-143 1.6SE-09 1.22E-06 1.3SE-10 No Data S.37E-10 No Data 4.S6E-OS Ce-144 4.BBE-07 2.04E-07 2.62E-OB No Data 1.21 E-07 No Data 1.6SE-04 Pr-143 9.20E-09 3.69E-09 4.S6E-10 No Data 2.13E-09 No Data 4.03E-OS Pr-144 3.01 E-11 1.2SE-11 1.S3E-12 No Data 7.05E-12 No Data 4.33E-18 Nd-147 6.29E-09 7.27E-09 4.3SE-10 No Data 4.2SE-09 No Data 3.49E-05 W-187 1.03E-07 B.61E-08 3.01E-OB No Data No Data No Data 2.82E-OS Np-239 1.19E-09 1.17E-10 6.4SE-11 No Data 3.6SE-10 No Data 2.40E-05 2-34 VER27
VEGPODCM Table 2-8. Site-Related Ingestion Dose Factors, Ai~
I Nuclide Bone Liver T.Body Thyroid Kidney Lung GI-LLI H-3 0.00 1.32E+00 1.32E+00 1.32E+00 1.32E+OO 1.32E+00 1.32E+OO C-14 3.13E+04 6.26E+03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 6.26E+03 Na-24 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 1.36E+02 P-32 1.32E+06 8.22E+04 5.11E+04 0.00 0.00 0.00 1.49E+05 Cr-51 0.00 0.00 1.27E+00 7.58E-Ol 2.79E-01 1.68E+00 3.19E+02 Mn-54 0.00 4.41E+03 8.42E+02 0.00 1.31E+03 0.00 1.35E+04 Mn-56 0.00 1.74E-01 3.08E-02 0.00 2.21 E-01 0.00 5.55E+00 Fe-55 6.86E+02 4.74E+02 1.11E+02 0.00 0.00 2.65E+02 2.72E+02 Fe-59 1.07E+03 2.51E+03 9.61E+02 0.00 0.00 7.01E+02 8.36E+03 Co-58 0.00 9.59E+01 2.15E+02 0.00 0.00 0.00 1.94E+03 Co-60 0.00 2.78E+02 6.14E+02 0.00 0.00 0.00 5.23E+03 Ni-63 3.25E+04 2.25E+03 1.09E+03 0.00 0.00 0.00 4.70E+02 Ni-65 1.72E-01 2.23E-02 1.02E-02 0.00 0.00 0.00 5.66E-01 Cu-64 0.00 2.75E+00 1.29E+00 0.00 6.94E+00 0.00 2.35E+02 Zn-65 2.32E+04 7.37E+04 3.33E+04 0.00 4.93E+04 0.00 4.64E+04 Zn-69 7.88E-07 1.51 E-06 1.05E-07 0.00 9.79E-07 0.00 2.26E-07 Br-83 0.00 0.00 3.83E-02 0.00 0.00 0.00 5.52E-02:
Br-84 0.00 0.00 1.22E-12 0.00 0.00 0.00 9.61 E-18 Br-85 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Rb-86 0.00 9.75E+04 4.54E+04 0.00 0.00 0.00 1.92E+04 Rb-88 0.00 1.29E-22 6.82E-23 0.00 0.00 0.00 1.78E-33 Rb-89 0.00 1.61E-26 1.14E-26 0.00 0.00 0.00 0.00 Sr-89 2.49E+04 0.00 7.16E+02 0.00 0.00 0.00 4.00E+03 Sr-90 6.23E+05 0.00 1.53E+05 0.00 0.00 0.00 1.80E+04 Sr-91 7.25E+01 0.00 2.93E+00 0.00 0.00 0.00 3.45E+02 Sr-92 3.33E-01 0.00 1.44E-02 0.00 0.00 0.00 6.60E+00 Y-90 5.04E-01 0.00 1.35E-02 0.00 0.00 0.00 5.34E+03 Y-91m 1.04E-11 0.00 4.01 E-13 0.00 0.00 0.00 3.04E-11 Y-91 9.77E+00 0.00 2.61 E-01 0.00 0.00 0.00 5.38E+03 Y-92 4.61E-04 0.00 1.35E-05 0.00 0.00 0.00 8.07E+00 Y-93 3.l9E-02 0.00 8.82E-04 0.00 0.00 0.00 1.01E+03 Zr-95 5.47E-Ol 1.75E-Ol 1.l9E-Ol 0.00 2.75E-01 0.00 5.56E+02 Zr-97 7.40E-03 1.49E-03 6.83E-04 0.00 2.26E-03 0.00 4.62E+02 Nb-95 8.09E+00 4.50E+00 2.42E+00 0.00 4.45E+00 0.00 2.73E+04 Mo-99 0.00 1.07E+02 2.04E+Ol 0.00 2.43E+02 0.00 2.49E+02 Tc-99m 5.70E-04 1.61E-03 2.05E-02 0.00 2.44E-02 7.89E-04 9.53E-01 All values are in (mrem*mL)/(h*JlCi). They are calculated using equation (2.14), and data from Table 2-5, Table 2-6, and Table 2-7. When "No Data" is shown for a radionuclide-organ combination in Table 2-7, Ait factors in this table are presented as zero.
2-35 VER27
VEGPODCM Table 2-8 (contd). Site-Related Ingestion Dose Factors, Air:
Nuclide Bone Liver T.Body Thyroid Kidney Lung GI-LLI Tc-101 2.71E-33 3.91E-33 3.83E-32 0.00 7.03E-32 2.00E-33 0.00.
Ru-103 6.21E+00 0.00 2.68E+00 0.00 2.37E+01 0.00 7.25E+021 Ru-105 8.79E-03 0.00 3.47E-03 0.00 1.14E-01 0.00 5.38E+001 Ru-106 9.42E+01 0.00 1.19E+01 0.00 1.82E+02 0.00 6.10E+03 Rh-105 2.32E+00 1.69E+00 1.11E+00 0.00 7.15E+00 0.00 2.68E+02 Ag-110m 2.53E+00 2.34E+00 1.39E+00 0.00 4.61E+00 0.00 9.56E+02 Sb-124 1.36E+03 2.56E+01 5.37E+02 3.28E+00 0.00 1.05E+03 3.84E+04 Sb-125 1.09E+03 1.17E+01 2.19E+02 9.68E-01 0.00 1.14E+05 9.63E+03 Te-125m 2.56E+03 9.29E+02 3.43E+02 7.71 E+02 1.04E+04 0.00 1.02E+04 Te-127m 6.51E+03 2.33E+03 7.93E+02 1.66E+03 2.64E+04 0.00 2.18E+04 Te-127 1.78E+01 6.40E+00 3.85E+00 1.32E+01 7.25E+01 0.00 1.41 E+03 Te-129m 1.09E+04 4.07E+03 1.73E+03 3.74E+03 4.55E+04 0.00 5.49E+04 Te-129 1.78E-05 6.68E-06 4.33E-06 1.36E-05 7.47E-05 0.00 1.34E-OS Te-131m 9.57E+02 4.68E+02 3.90E+02 7.42E+02 4. 74E+03 0.00 4.65E+04 Te-131 8.64E-17 3.61 E-17 2.73E-17 7.10E-17 3.78E-16 0.00 1.22E-17 Te-132 1.97E+03 1.27E+03 1.19E+03 1.41 E+03 1.23E+04 0.00 6.02E+04 1-130 7.60E+00 2.24E+01 8.85E+00 1.90E+03 3.50E+01 0.00 1.93E+01*
1-131 1.73E+02 2.48E+02 1.42E+02 8.13E+04 4.2SE+02 0.00 6.S5E+01 1-132 S.27E-03 1.41 E-02 4.93E-03 4.93E-01 2.24E-02 0.00 2.65E-03 1-133 2.S9E+01 4.S1E+01 1.37E+01 6.62E+03 7.86E+01 0.00 4.0SE+01 1-134 2.18E-08 5.94E-08 2.12E-08 1.03E-06 9.44E-08 0.00 5.17E-11 1-13S 1.31E+00 3.44E+00 1.27E+00 2.27E+02 S.52E+00 0.00 3.89E+00 Cs-134 2.98E+OS 7.10E+OS 5.80E+OS 0.00 2.30E+OS 7.62E+04 1.24E+04 Cs-136 2.96E+04 1.17E+OS 8.42E+04 0.00 6.S1E+04 8.92E+03 1.33E+04 Cs-137 3.82E+05 S.23E+05 3.43E+OS 0.00 1.78E+OS 5.90E+04 1.01E+04 Cs-138 9.12E-12 1.80E-11 8.92E-12 0.00 1.32E-11 1.31E-12 7.68E-17 Ba-139 S.64E-06 4.02E-09 1.65E-07 0.00 3.76E-09 2.28E-09 1.00E-OS Ba-140 3.74E+02 4. 69E-01 2.45E+01 0.00 1.60E-01 2.69E-01 7.69E+02 Ba-141 8.47E-25 6.40E-28 2.86E-26 0.00 S.9SE-28 3.63E-28 3.99E-34 Ba-142 0.00 0.00 0.00 0.00 0.00 0.00 0.00 La-140 1.10E-01 5. 56E-02 1.47E-02 0.00 0.00 0.00 4.08E+03 La-142 2.19E-07 9.96E-08 2.48E-08 0.00 0.00 0.00 7.27E-04 Ce-141 1.15E-01 7. 79E-02 8.B4E-03 0.00 3.62E-02 0.00 2.98E+02 Ce-143 8.6SE-03 6.39E+00 7.08E-04 0.00 2.81E-03 0.00 2.39E+02 Ce-144 6.22E+00 2.60E+00 3.34E-01 0.00 1.54E+00 0.00 2.10E+03 Pr-143 6.10E-01 2.44E-01 3.02E-02 0.00 1.41 E-01 0.00 2.67E+03 Pr-144 1.48E-28 6.14E-29 7.S1 E-30 0.00 3.46E-29 0.00 2.13E-3S Nd-147 4.11 E-01 4.7SE-01 2.B4E-02 0.00 2.78E-01 0.00 2.28E+03 W-187 1.47E+02 1.23E+02 4.31E+01 0.00 0.00 0.00 4.04E+04 Np-239 2.81E-02 2.76E-03 1.52E-03 0.00 8.62E-03 0.00 S.67E 2-36 VER27
VEGPODCM 2.5 LIQUID EFFLUENT DOSE PROJECTIONS 2.5.1 Thirty-One Day Dose Projections In order to meet the requirements for operation of the LIQUID RADWASTE TREATMENT SYSTEM (see Section 2.1.4), dose projections must be made at least once each 31 days; this applies during periods in which a discharge to UNRESTRICTED AREAS of liquid effluents containing radioactive materials occurs or is expected.
Projected 31-day doses to individuals due to liquid effluents may be determined as follows:
(2.17) where:
D'tp = the projected dose to the total body or organ 1:, for the next 31 days of liquid releases.
Dw = the cumulative dose to the total body or organ 1:, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.
t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter).
D-ra = the anticipated dose contribution to the total body or any organ 1:, due to any planned activities during the next 31-day period, if those activities will result in liquid releases that are in addition to routine liquid effluents. If only routine liquid effluents are anticipated, D-ra may be set to zero.
2.5.2 Dose Projections for Specific Releases Dose projections may be performed for a particular release by performing a prerelease dose calculation assuming that the planned release will proceed as anticipated. For individual dose projections due to liquid releases, follow the methodology of Section 2.4, using sample analysis results for the source to be released, and parameter values expected to exist during the release period.
2-37 VER27
VEGPODCM 2.6 DEFINITIONS OF LIQUID EFFLUENT TERMS The following symbolic terms are used in the presentation of liquid effluent calculations in the subsections above.
Section of Definition Initial Use the adjustment factor used in calculating the effluent monitor setpoint for liquid release pathway p: the ratio of the assured dilution to the required dilution [unitless]. 2.3.2.2 ADF= the assured dilution factor for a planned release [unitless].
I 2.3.2.2 AFp= the dilution allocation factor for liquid release pathway p 2.3.2.2
[unitless]. I the site-related adult ingestion dose commitment factor, for the total body or for any organ t, due to identified radionuclide i [(mrem . mL)/(h . j.1Ci)]. The values of Ai't are listed in Table 2i8. 2.4.1 the crop to soil concentration factor applicable to radionuclide i, [(pCilkg garden vegetation)/(pCi/kg soil)]. 2.4.3 the bioaccllmulation factor for radionuclide i for freshwater fish [(pCilkg)/(pCilL)]. Values are listed in Table 2-6. 2.4.2 c= the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line, prior to dilution and subsequent release [/J.Ci/ml]. 2.3.2.1 the calculated effluent radioactivity monitor setpoint for liquid release pathway p [/J.Ci/ml]. 2.3.2.2 Ca = the gross concentration of alpha emitters in the liquid waste as measured in the applicable composite sample [/J.Ci/ml]. 2.3.2.2 CECL= the Effluent Concentration limit stated in 10 CFR 20. Appendix B, Table 2, Colunin 2 [/J.Ci/ml]. 2.3.2.1 the concentration of Fe-55 in the liquid waste as measured in the applicable composite sample [IlCi/mL]. 2.3.2.2 2-38 VER27
VEGPODCM Section of Term Definition Initial Use Cg = the concentration of gamma emitter g in the liquid waste as measured by gamma ray spectroscopy performed on the applicable prerelease waste sample [)lCi/mL]. 2.3.2.2 Ci= the measured concentration of radionuclide i in a sample of liquid effluent [)lCilmL]. 2.3.2.2 CiI= the average concentration of radionuclide i in undiluted liquid effluent during time period I [)lCi/mL]. 2.4.1 Cir = the measured concentration of radionuclide i in release pathway r that is contributing to radioactivity in the dilution 2.3.2.2 stream [)lCi/mL].
Cs = the concentration of strontium radioisotope s (Sr-89 or Sr-90) in the liquid waste as measured in the applicable composite sample [)lCi/mL]. 2.3.2.2 Ct = the concentration of H-3 in the liquid waste as measured in the applicable composite sample [)lCi/mL]. 2.3.2.2 CFiv = the concentration factor for radionuclide i in irrigated garden vegetation [(pCi/kg)/(pCilL)]. 2.4.2 Dw= the dilution factor from the near field of the discharge structure to the potable water intake location [unitless]. 2.4.2 D,;= the cumulative dose commitment to the total body or to any organ 1:, due to radioactivity in liquid effluents released during a given time period [mrem]. 2.4.1 D,;a = the anticipated dose contribution to the total body or any organ 1:, due to any planned activities during the next 31-day period [mrem]. 2.5.1 D'tC = the cumulative dose to the total body or organ 't, for liquid releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration 2.5.1
[mrem].
D1:p = the projected dose to the total body or organ 't, for the next 31 days of liquid releases [mrem]. 2.5.1 2-39 VER27
VEGPODCM Section of Term Definition Initial Use DFi't= the dose conversion factor for radionuclide i for adults, in organ 1: [mrem/pCi]. Values are listed in Table 2-7. 2.4.2 EC~= the liquid Effluent Concentration Limit for radionuclide i from 10 CFR Part 20, Appendix B. Table 2, Column 2 [J-lCi/mL]. 2.3.2.2 f= the effluent flowrate at the location of the radioactivity monitor [gpm]. 2.3.2.1 fap = the anticipated actual discharge flowrate for a planned release from liquid release pathway p [gpm]. 2.3.2.2 f/ = the fraction of the year that garden vegetation is irrigated [unitless]. 2.4.3 fmp = the maximum permissible effluent discharge flowrate for release pathway p [gpm). 2.3.2.2 fr = the effluent discharge flowrate of release pathway r [gpm]. 2.3.2.2 ft = the average undiluted liquid waste flowrate actually observed during the period of a liquid release [gpm]. 2.4.1 F= the dilution stream flowrate which can be assured prior to the release point to the UNRESTRICTED AREA [gpm). 2.3.2.1 Fd = the entire assured dilution flowrate for the plant site during the release period [gpm]. 2.3.2.2 Fdp= the dilution flowrate allocated to release pathway p [gprn). 2.3.2.2 F,= the near-field average dilution factor in the receiving water of the UNRESTRICTED AREA [unitless]. 2.4.1 Ft = the average dilution stream flowrate actually observed during the period of a liquid release [gpm]. 2.4.1 1= the average irrigation rate during the growing season [U(m2 2.4.3
. h>>).
Lv = the water content of leafy garden vegetation edible parts 2.4.3
[Ukg].
2-40 VER27
VEGPODCM Section of Term Definition Initial Use M== the additional river dilution factor from the near field of the discharge structure for the plant site to the pOint of irrigation water usage [unitless]. 2.4.3 p= the effective surface density of soil [kg/m 2]. 2.4.3 r= the fraction of irrigation*deposited activity retained on the edible portions of leafy garden vegetation. 2.4.3 RDF= the required dilution factor: the minimum ratio by which liquid effluent must be diluted before reaching the UNRESTRICTED AREA, in order to ensure that the limits of Section 2.1.2 are not exceeded [unitless]. 2.3.2.2 RDFy = the RDF for a liquid release due only to its concentration of gamma*emitting radionuclides [unitless]. 2.3.2.2 RDFny = the RDF for a liquid release due only to its concentration of non*gamma*emitting radionuclides [unitless]. 2.3.2.2 SF == the safety factor selected to compensate for statistical fluctuations and errors of measurement [unitless]. 2.3.2.2 t= the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration. 2.5.1 itJ= the period of long*term buildup of activity in soil [h]. 2.4.3 te = the period of leafy garden vegetation exposure during the growing season [h]. 2.4.3 tf = the transit time from release to receptor for fish 2.4.2 consumption [h].
th = the time between harvest of garden vegetation and human consumption [h]. 2.4.3 tw = the transit time from release to receptor for potable water consumption [h]. 2.4.2 2-41 VER27
VEGPODCM Section of Term Definition Initial Use TF= the tolerance factor selected to allow flexibility in the establishment of a practical monitor setpoint which could accommodate effluent releases at concentrations higher than the ECl values stated in 10 CFR 20, Appendix B, Table 2, Column 2 [unitless]; the tolerance factor must not 2.3.2.1 exceed a value of 10.
Uf= the adult rate of fish consumption [kgly]. 2.4.2 Uv = the adult consumption rate for irrigated garden vegetation [kgly]. 2.4.2 Uw= the adult drinking water consumption rate applicable to the plant site [Uy]. 2.4.2 Yv= the areal density (agricultural productivity) of leafy garden vegetation [kglm 2]. 2.4.3 z= the applicable dilution factor for the receiving water body, in the near 'field of the discharge structure, during the period of radioactivity release [unitless]. 2.4.1
.1t1 = the length of time period 1, over which Ci1 and F1 are averaged for liquid releases [h]. 2.4.1 AEi= the effective removal rate for activity deposited on crop leaves [h-1]. 2.4.3 Ai= the decay constant for radionuclide i [h- 1]. 2.4.2 Aw= the rate constant for removal of activity from plant leaves by weathering [h- 1]. 2.4.3 2-42 VER27
VEGPODCM CHAPTER 3 GASEOUS EFFLUENTS 3.1 LIMITS OF OPERATION The following Limits of Operation implement requirements established by Technical Specifications Section 5.0. Terms printed in all capital letters are defined in Chapter 10.
3.1.1 Gaseous Effluent Monitoring Instrumentation Control In accordance with Technical Specification 5.5.4.a, the radioactive gaseous effluent monitoring instrumentation channels shown in Table 3-1 shall be OPERABLE with their alarm/trip setpoints set to ensure that the limits of Section 3.1.2.a are not exceeded. The alarm/trip setpoints of these channels shall be determined in accordance with Section 3.3.
3.1.1.1 Applicability These limits apply as shown in Table 3-1.
3.1 .1.2 Actions With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above control, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, declare the channel inoperable, or restore the setpoint to a value that will ensure that the limits of Section 3.1 .2.a are met.
With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3-1. Restore the inoperable instrumentation to operable status within 30 days, or if unsuccessful, explain in the next Radioactive Effluent Release Report, per Technical Specification 5.6.3, why this inoperability was not corrected in a timely manner.
This control does not affect shutdown requirements or MODE changes.
3.1 .1.3 Surveillance Requirements Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 3-2.
3-1 VER27
VEGP ODCM 3.1.1 .4 Basis The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actua.l or potential releases of gaseous effluents. The AlarmfTrip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in Section 3.3 to ensure that the alarm/trip will occur prior to exceeding the limits of Section 3.1.2.a. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.
3-2 VER27
VEGPODCM Table 3-1. Radioactive Gaseous Effluent Monitoring Instrumentation OPERABILITY Requirements Minimum Instrument Channels OPERABLE Applicability ACTION
1. GASEOUS RADWASTE TREATMENT SYSTEM (Common)
a. Noble Gas Activity Monitor, with Alarm and Automatic Termination of Release (ARE-0014) 1 During releasesa 45
b. Effluent System Flowrate Measuring Device (AFT-0014) 1 During releasesa 46
2. Turbine Building Vent (Each Unit)
a. Noble Gas Activity Monitor (RE-12839C) 1 During releases a 47
b. Iodine and Particulate Samplers (RE-12839A & B) 1 During releasesa 51
c. Flowrate Monitor (FT-12839 or FIS-12862)b 1 During releasesa 46
d. Sampler Flowrate Monitor (1FI-13211,2FIT-13211) 1 During releasesa 46
3. Plant Vent (Each Unit)
a. Noble Gas Activity Monitor (RE-12442C or RE-12444C) 1 At all times 47,48
b. Iodine Sampler/Monitor (RE-12442B or RE-12444B) 1 At all times 51
c. Particulate Sampler/Monitor (RE-12442A or RE-12444A) 1 At all times 51
d. Flowrate Monitor (FT-12442 or 12835) 1 At all times 46
e. Sampler Flowrate Monitor (FI-12442 or FI-12444) 1 At all times 46
4. Radwaste Processing Facility Vent (Common)
a. Particulate Monitor (ARE-16980) 1 During releasesa 51
a. "During releases" means "During radioactive releases via this pathway."
b. During emergency filtration.
3-3 VER27
VEGPODCM Table 3*1 (contd). Notation for Table 3*1 - ACTION Statements ACTION 45 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank(s) may be released to the environment for up to 14 days provided that prior to initiating the release:
a. The local radiation monitor reading (if functional) is recorded at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or at least two independent samples of the tank's contents are analyzed, and
b. At least two technically qualified members of the Facility Staff independently verify the discharge line valving, and verify the release rate calculations.
Otherwise, suspend release of radioactive effluents via this pathway.
ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flowrate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
ACTION 47 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the local radiation monitor reading (if functional) is recorded at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or grab samples are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
ACTION 48 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, record the local radiation monitor reading (if functional) at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or immediately suspend containment purging of radioactive effluents via this pathway.
ACTION 49 - (Not Used)
ACTION 50 - (Not Used)
ACTION 51 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided the local radiation monitor reading (if functional) is recorded at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or samples are continuously collected with auxiliary sampling equipment. RE-12444A and B may be verified functional by recording local radiation monitor skid flow once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
3-4 VER 27
VEGPODCM Table 3-2. Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements Surveillance Requirements CHANNEL CHANNEL OPERA-Instrument CHANNEL SOURCE CALI BRA- TIONAL CHECK CHECK TION TEST MODESc
1. GASEOUS RADWASTE TREATMENT SYSTEM (Common)
a. Noble Gas Activity Monitor, with Alarm and Automatic Termination of Release During (ARE-0014) P P Rb Ra (1) Release
b. Effluent System Flowrate Measuring Device During (AFT-0014) P NA R NA Release
2. Turbine Building Vent (Each Unit)
a. Noble Gas Activity Monitor During (RE-12839C) D M Rb Ra(2) Release
b. Iodine and Particulate During Samplers (RE-12839A&B) Wd NA NA NA Release
c. Flowrate Monitor During (FT-12839 or FIS-12862) D NA R NA Release
d. Sampler Flowrate Monitor During (1 FI-13211, 2FIT-13211) D NA R Q Release
3. Plant Vent (Each Unit)
a. Noble Gas Activity Monitor (RE-12442C or RE-12444C) D M Rb Ra (2) All
b. Particulate and Iodine Mon itors (RE-12442A&B) Wd NA R Ra(2) All
c. Particulate and Iodine Samplers (RE-12444A&B) Wd NA NA NA All
d. Flowrate Monitor (FT-12442 or 12835) D NA R NA All
e. Sampler Flowrate Monitor (FI-12442 or FI-12444) D NA R Q I All
4. Radwaste Processing Facility Vent (Common)
a. Particulate Monitor During Wd Q RS N/A (ARE-16980) Release 3-5 VER27
VEGPODCM Table 3-2 (contd). Notation for Table 3-2
a. In addition to the basic functions of a CHANNEL OPERATIONAL TEST (Section 10.2):
(1) The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room annunciation occurs (for item a. below only); and control room CRT indication occurs (if any of the following conditions exist):
(a) Instrument indicates measured levels above the alarmltrip setpoint; (b) Instrument indicates an "Equipment Trouble" alarm; (c) Instrument indicates a "Low" alarm; or (d) Instrument indicates channel "Deactivated."
(2) The CHANNEL OPERATIONAL TEST shall also demonstrate that control room annunciation occurs (for item a. below only); and that control room CRT indication occurs (if any of the following conditions exist):
(a) Instrument indicates measured levels above the alarm/trip setpoint; (b) Instrument indicates an "Equipment Trouble" alarm; (c) Instrument indicates a "Low" alarm; or (d) Instrument indicates channel "Deactivated." ("Loss of counts" for ARE 16980 only)
b. The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology, or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST. These standards shall permit calibrating the system over its intended range of energy and measurement range. For any subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.
c. MODES in which surveillance is required. "All" means "At all times." "During release" means "During radioactive release via this pathway."
d. The channel check shall consist of visually verifying that the collection device (Le.,
particulate filter or charcoal cartridge, etc.) is in place for sampling.
e. The CHANNEL CALIBRATION verifies proper operation of the CHANNEL OPERATIONAL TEST requirements described in Notation a(2) above.
3-6 VER27
VEGP ODCM 3.1 .2 Gaseous Effluent Dose Rate Control In accordance with Technical Specifications 5.5.4.c and 5.5.4.g, the licensee shall conduct operations so that the dose rates due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY are limited as follows:
a. For noble gases: Less than or equal to a dose rate of 500 mrem/y to the total body and less than or equal to a dose rate of 3000 mrem/y to the skin, and
b. For lodine-131, lodine-133, tritium, and for all radionuclides in particulate form with half lives greater than 8 days: Less than or equal to a dose rate of 1500 mrem/y to any organ.
3.1.2.1 Applicability This limit applies at all times.
3.1.2.2 Actions With a dose rate due to radioactive material released in gaseous effluents exceeding the limit stated in Section 3.1.2, immediately decrease the release rate to within the stated limit.
These limits do not affect shutdown requirements or MODE changes.
3.1.2.3 Surveillance Requirements The dose rates due to radioactive materials in areas at or beyond the SITE BOUNDARY due to releases of gaseous effluents shall be determined to be within the above limits, in accordance with the methods and procedures in Section 3.4.1 , by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 3-3.
3.1.2.4 Basis This control is provided to ensure that gaseous effluent dose rates will be maintained within the limits that historically have provided reasonable assurance that radioactive material discharged in gaseous effluents will not result in a dose to a MEMBER OF THE PUBLIC in an UNRESTRICTED AREA, either within or outside the SITE BOUNDARY, exceeding the limits specified in Appendix I of 10 CFR Part 50, while allowing operational flexibility for effluent releases. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of the MEMBER OF THE PUBLIC will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the SITE BOUNDARY.
The dose rate limit for lodine-131, lodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days specifically applies to dose rates to a child via the inhalation pathway.
This control applies to the release of gaseous effluents from all reactors at the site.
3-7 VER27
VEGPODCM Table 3-3. Radioactive Gaseous Waste Sampling and Analysis Program Sampling and Analysis Requirements a MINIMUM DETECTABLE Minimum CONCENTRA-Gaseous Sampling Analysis Type of Activity TION (MDC)
Release Type FREQUENCY FREQUENCY Analysis (J.lCilmL)
Waste Gas P P Noble Gas 1 E-4 Decay Tank Each Tank Grab Each Tank PRINCIPAL (Common) Sample GAMMA EMITIERS pc Noble Gas 1 E-4 Containment pc Purge Each Purge PRINCIPAL Each Purge GAMMA EMITTERS 24" or 14" (Each Unit) Grab Sample M H-3 (Oxide) 1 E-6 Noble Gas 1 E-4 Mc,d,! PRINCIPAL Plant Vent MC (Each Unit) GAMMA EMITTERS Grab Sample H-3 (Oxide) 1 E-6 Noble Gas 1 E-4 Condenser Air PRINCIPAL Ejector &
M GAMMA EMITTERS Steam Packing M Grab Sample Exhaust (Each Unit)b H-3 (Oxide) 1 E-6 We 1-131 1 E-12 CONTINUOUS9 Charcoal or Silver Zeolite Sample We Particulate 1 E-11 CONTINUOUS9 Particulate PRINCIPAL Sample GAMMA EMITTERS Plant Vent, M Gross Alpha 1 E-11 Condenser Air COMPOSITE Ejector & Particulate Steam Packing CONTINUOUS9 Sample Exhaust (Each Unit)b Q Sr-89, Sr-90 1 E-11 CONTINUOUS9 COMPOSITE Particulate Sample Radwaste Wh Particulate 1 E-11 Processing Particulate PRINCIPAL I Facility Vent CONTINUOUS9 Sample GAMMA EMITTERS
(Common) 3-8 VER27
VEGPODCM Table 3-3 (contd). Notation for Table 3-3
a. Terms printed in all capital letters are defined in Chapter 10.
b. The turbine building vent is the release point for the condenser air ejector and steam packing exhaust. All sampling and analyses may be omitted for this vent, provided the absence of a primary to secondary leak has been demonstrated, that is, if the gamma activity in the secondary water does not exceed background by more than 20%.
c. Sampling and analysis shall also be performed following shutdown, startup, or a THERMAL POWER change exceeding 15% of the RATED THERMAL POWER within a one-hour period. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
d. Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the refueling cavity is flooded.
e. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler). Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each shutdown, startup, or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in one hour, and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding MDC may be increased by a factor of 10. This requirement does not apply if (1) analysis shows that the DOSE EQUIVALENT 1-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that effluent activity has not increased more than a factor of 3.
f. Tritium grab samples shall be taken at least once per 7 days from the Unit 1 plant vent, whenever spent fuel is in the spent fuel pool (Unit 1 plant vent only).
g. The ratio of the sample flowrate to the sampled stream flowrate shall be known for the time period covered by each dose or dose rate calculation made in accordance with controls specified in Sections 3.1.2, 3.1.3, and 3.1.4.
h. Samples shall be changed at least once per 7 days and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or removal of sampler).
3-9 VER27
VEGPODCM 3.1.3 Gaseous Effluent Air Dose Control In accordance with Technical Specifications 5.5.4.e and 5.5.4.h, the air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNDARY shall be limited to the following:
a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and
b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
3.1.3.1 Applicability This limit applies at all times.
3.1.3.2 Actions With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days a special report which identifies the cause(s) for exceeding the limit(s); defines the corrective actions that have been taken to reduce the releases; and defines the proposed corrective actions to be taken to assure that subsequent releases of radioactive noble gases in gaseous effluents will be in compliance with the limits of Section 3.1.3.
This control does not affect shutdown requirements or MODE changes.
3.1.3.3 Surveillance Requirements Cumulative air dose contributions from noble gas radionuclides released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.2 at least once per 31 days.
3.1.3.4 Basis This control is provided to implement the requirements of Sections II.B, III.A and IV.A of Appendix I, 10 CFR Part 50. Section 3.1.3 implements the guides set forth in Section II.B of Appendix I.
The ACTION statements in Section 3.1.3.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I, assuring that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The Surveillance requirements in Section 3.1.3.3 implement the requirements in Section III.A of Appendix I, which require that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in Section 3.4.2 for calculating the doses due to the actual releases of noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). The equations in Section 3.4.2 provided for determining the air doses at the SITE BOUNDARY are based upon the historical annual average atmospheric conditions.
3-10 VER27
VEGPODCM 3.1.4 Control on Gaseous Effluent Dose to a Member of the Public In accordance with Technical Specifications 5.5.4.e and 5.5.4.i, the dose to a MEMBER OF THE PUBLIC from 1-131,1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each reactor unit, to areas at and beyond the SITE BOUNDARY shall be limited to the following:
a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ, and
b. During any calendar year: Less than or equal to 15 mrem to any organ.
3.1.4.1 Applicability This limit applies at all times.
3.1.4.2 Actions With the calculated dose from the release of 1-131, 1-133, tritium, or radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Nuclear Regulatory Commission within 30 days a special report which identifies the cause(s) for exceeding the limit; defines the corrective actions that have been taken to reduce the releases of radioiodines and radionuclides in particulate form with half-lives greater than 8 days in gaseous emuents; and defines proposed corrective actions to assure that subsequent releases will be in compliance with the limits stated in Section 3.1.4.
This control does not affect shutdown requirements or MODE changes.
3.1.4.3 Surveillance Requirements Cumulative organ dose contributions to a MEMBER OF THE PUBLIC from 1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days released in gaseous effluents from each unit to areas at and beyond the SITE BOUNDARY, for the current calendar quarter and current calendar year, shall be determined in accordance with Section 3.4.3 at least once per 31 days.
3.1 .4.4 Basis This control is provided to implement the requirements of Section II.C, liLA and IV.A of Appendix I, 10 CFR Part 50. The limits stated in Section 3.1.4 are the guides set forth in Section II.C of Appendix I. The ACTION statements in Section 3.1.4.2 provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." The calculational methods specified in the Surveillance Requirements of Section 3.1.4.3 implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The calculational methods in Section 3.4.3 for calculating the doses due to the actual releases of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109 (Reference 3), and Regulatory Guide 1.111 (Reference 5). These equations provide for determining the actual doses 3-11 VER27
VEGPODCM based upon the historical annual average atmospheric conditions. The release specifications for radioiodines, radioactive materials in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the areas at and beyond the SITE BOUNDARY. The pathways which were examined in the development of these calculations were: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides onto green leafy garden vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk animals and meat producing animals graze with consumption of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man.
3.1.5 Gaseous Radwaste Treatment System Control In accordance with Technical Specification 5.5.4.f, the GASEOUS WASTE PROCESSING SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS WASTE PROCESSING SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous wastes prior to their discharge when the projected doses in 31 days due to gaseous effluent releases, from each reactor unit, to areas at and beyond the SITE BOUNDAR" would exceed 0.2 mrad to air from gamma radiation, 0.4 mrad to air from beta radiation, or 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.
3.1.5.1 Applicability These limits apply at all times.
3.1.5.2 Actions With gaseous waste being discharged without treatment and in excess of the limits in Section 3.1.5, prepare and submit to the Nuclear Regulatory CommiSSion within 30 days a special report which includes the following information:
a. Identification of any inoperable equipment or subsystem and the reason for inoperability,
b. Action(s) taken to restore the inoperable equipment to OPERABLE status, and
c. Summary description of action(s) taken to prevent a recurrence.
This control does not affect shutdown requirements or MODE changes.
3.1.5.3 Surveillance Requirements Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days, in accordance with Section 3.5.1 ,when the GASEOUS WASTE PROCESSING SYSTEM or the VENTILATION EXHAUST TREATMENT SYSTEM is not being fully utilized.
The GASEOUS WASTE PROCESSING SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be demonstrated OPERABLE:
3-12 VER27
VEGPOOCM by meeting the controls of Sections 3.1.2, and either 3.1.3 (for the GASEOUS WASTE PROCESSING SYSTEM) or 3.1.4 (for the VENTILATION EXHAUST TREATMENT SYSTEM).
3.1.5.4 Basis The OPERABILITY of the GASEOUS WASTE PROCESSING 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 control implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section 11.0 of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the system were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents.
This control applies to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared radwaste treatment systems, the gaseous effluents from the shared system are proportioned among the units sharing that system.
3.1.6 Major Changes to Gaseous Radioactive Waste Treatment Systems Licensee initiated MAJOR CHANGES TO GASEOUS RADIOACTIVE WASTE TREATMENT SYSTEMS:
a. Shall be reported to the Nuclear Regulatory Commission in the Radioactive Effluents Release Report for the period in which the change was implemented. The discussion of each change shall contain the information described in Section 7.2.2.7.
b. Shall become effective upon review and approval by the Vice President Plant.
3-13 VER27
VEGPODCM 3.2 GASEOUS WASTE PROCESSING SYSTEM At Plant Vogtle, there are five potential paints where radioactivity may be released to the atmosphere in gaseous discharges. These five potential release pathways are the Unit 1 and Unit 2 Plant Vents; the Unit 1 and Unit 2 Turbine Building Vents; and the Radwaste Processing Facility Vent. However, the Turbine Building Vents are not normal release pathways unless a primary-to-secondary leak exists. The Radwaste Processing Facility Vent is not a normal release pathway unless a spill occurs. The figures on the following pages give schematic diagrams of the Gaseous Waste Treatment System and the Ventilation Exhaust Treatment Systems (Reference 11 ).
The Unit 1 Plant Vent release pathway includes two release sources that are common to the two units: ventilation air from the Fuel Handling Building, and discharges from the GASEOUS WASTE PROCESSING SYSTEM. Otherwise, discharges from the two reactor units are separated. Reactor Containment Building ventilation releases are through the respective plant vents. The Turbine Building Vent serves as the discharge point for both the condenser air ejector and the steam packing exhauster system. The Radwaste Processing Facility Vent includes sources from the Radwaste Processing Facility Process area.
Releases from the two Turbine Building Vents and the Radwaste Processing Facility Vent are considered to be ground-level releases, whereas releases from the two Plant Vents are considered mixed-mode releases. Chapter 8 discusses the calculation of atmospheric dispersion parameters using the ground-level and mixed-mode (I.e., split-wake) models. All five potential release pathways are considered to be continuous (as opposed to batch) in nature.
3-14 VER27
VEGPODCM Radioactivity Monitor ARE*0014 ~ To Unit 1 Plant Vent
,............- - via Auxiliary Building Ventilation System Catalytic Recombiner and Waste Gas Compressor Gas Analyzer Package 1-.......__ To Chemical Volume Control Tank I
I
I To Waste Gas Decay I I I 1-......- - Tank Header I------.t}---~
Waste Gas Decay Tank Waste Gas Decay Tank (Shutdown)
(Seven per Unit)
(Two Shared)
From Unit 2 Waste Gas Decay Tanks Volume Control Tank Purge Recycle Evaporator Vent Condenser Waste Evaporator Vent Condenser Recycle Holdup Tank Eductor Reactor Coolant Drain Tank
'Dotted line operational between 20 and 100 psig NOTE: This is typical of both units. However, Unit 2 GWPS releases via Unit 1 plant vent.
Figure 3-1. Schematic Diagram of the Gaseous Radwaste Treatment System 3-15 VER27
VEGPODCM Plant Vent Radioactivity Monitor 1RE12442A.B.C t
HEPA HEPA HEPA CF CF CF HEPA HEPA HEPA HC HC Radioactivit Y Monitor R 0039A ME ME Auxiliary Building Radioactivity Monitor 1RE2565A.B.C Fuel Handl ing Building (Shared)""
Radioactivit y Monitor Reactor Containment ARE-0014 From Waste Gas Processing Area and Svstem HEPA - High-Efficiency Particulate Air Filter CF - Activated Charcoal Filter HC - Heating Coil ME - Moisture Eliminator
"" Prior to treatment by the Fuel Handling Building Ventilation Exhaust Treatment System, Exhaust from Unit 1 Spent Fuel Pool Area is monitored by ARE2532B and ARE2533B; exhaust from Unit 2 Spent Fuel Pool Area is monitored by ARE2532A and ARE2533A.
Figure 3-2. Schematic Diagram of the Unit 1 Plant Vent Release Pathway 3-16 VER27
VEGPODCM Plant Vent Radioactivity Monitor 2RE12442A,B,C I
HEPA HEPA I
I CF CF I
HEPA HEPA I HC Auxiliary Building ME Radioactivity Mnnitor Reactor Containment HEPA - High-Efficiency Particulate Air Filter CF - Activated Charcoal Filter HC - Heating Coil ME - Moisture Eliminator Figure 3-3. Schematic Diagram of the Unit 2 Plant Vent Release Pathway 3-17 VER27
VEGPODCM Turbine Building Vent f
9 Radioactivity Monitor 1(2)RE12839A,B,C
~
J HEPA HEPA I
CF CF HEPA HEPA HC HC DE DE
,- NO - NC _NO - NC Steam Jet Air Steam Packing HEPA - High-Efficiency Particulate Air Filter CF - Activated Charcoal Filter HC - Heating Coil DE - Demister NO - Normally Open NC - Normally Closed NOTE: This is typical of both units.
Figure 3-4. Schematic Diagram of the Turbine Building Vent Release Pathway (Typical of Both Units) 3-18 VER27
VEGPODCM Radiollctivity MCIIIJoit<< ARE 16980 HBPA Figure 3-5. Schematic Diagram of the Radwaste Processing Facility Ventilation Release Pathway 3-19 VER27
VEGPODCM 3.3 GASEOUS EFFLUENT MONITOR SETPOINTS 3.3.1 General Provisions Regarding Noble Gas Monitor Setpoints Noble gas radioactivity monitor setpoints calculated in accordance with the methodology presented in this section are intended to ensure that the limits of Section 3.1.2.a are not exceeded. They will be regarded as upper bounds for the actual high alarm setpoints. That is, a lower high alarm setpoint may be established or retained on the monitor, if desired. Intermediate level setpoints should be established at an appropriate level to give sufficient warning prior to reaching the high alarm setpoint.
If no release is planned for a given pathway, or if there is no detectable activity in the gaseous stream being evaluated for release, the setpoint should be calculated in accordance with the methods presented below, based on an assumed concentration of Kr-88 that leads to a practical setpoint. A practical setpoint in this context is one which prevents spurious alarms, and yet produces an alarm should a significant inadvertent release occur.
Section 3.1.1 establishes the requirements for gaseous effluent monitoring instrumentation, and Section 3.2 describes the VENTILATION EXHAUST TREATMENT SYSTEM and the GASEOUS WASTE PROCESSING SYSTEM. From those Sections, it can be seen that certain monitors are located on final release pathways, that is, streams that are being monitored immediately before being discharged from the plant; the setpoint methodology for these monitors is presented in Section 3.3.2. Other monitors are located on source streams, that is, streams that merge with other streams prior to passing a final monitor and being discharged; the setpoint methodology for these monitors is presented in Section 3.3.3. Table 3-4 identifies which of these setpoint methodologies applies to each monitor. Some additional monitors with special setpoint requirements are discussed in Section 3.3.5.
As established in Section 3.1.1 , gaseous effluent monitor setpoints are required only for the noble gas monitors on certain potential release streams: the two Plant Vents, the two Turbine Building Vents, and the GASEOUS WASTE PROCESSING SYSTEM discharge. However, because of the potential significance of releases from other sources, Section 3.3 discusses setpoint methodologies for certain additional monitors, as well.
3-20 VER27
VEGPODCM Table 3-4. Applicability of Gaseous Monitor Setpoint Methodologies Final Release Pathways with no Monitored Source Streams Setpoint Method: Section 3.3.2 Release Elevation: Ground-level Unit 1 or Unit 2 Turbine Building Vent Monitor: 1RE-12839C/2RE-12839C Maximum Flowrate: 900 cfm (4.25 E+05 mUs)
Final Release Pathways with One or More Monitored Source Streams Release Elevation: Mixed-Mode Unit 1 Plant Vent Monitors: 1RE-12442C,1RE-12444C Maximum Flowrate: 187,000 cfm (8.83 E+07 mUs)
Setpoint Method: Section 3.3.2 Release Type: CONTINUOUS Source Stream: Unit 1 Reactor Containment Purge Monitor: 1RE-2565C Maximum Flowrate: release-dependent Setpoint Method: Section 3.3.3 Release Type: BATCH Source Stream: Gaseous Waste Treatment System Monitor: ARE-0014 Maximum Flowrate: release-dependent Setpoint Method: Section 3.3.3 Release Type: BATCH Unit 2 Plant Vent Monitors: 2RE-12442C, 2RE-12444C Maximum Flowrate: 112,500 cfm (5.31 E+07 mUs)
Setpoint Method: Section 3.3.2 Release Type: CONTINUOUS Source Stream: Unit 2 Reactor Containment Purge Monitor: 2RE-2565C Maximum Flowrate: release-dependent Setpoint Method: Section 3.3.3 Release Type: BATCH (X7afvb Values for Use in Setpoint Calculations Ground-Level Releases: 2.55 x 10-6 s/m 3 [NE Sector]
Mixed-Mode Releases: 4.62 x 10-7 s/m 3 [NE Sector]
Maximum flowrate values are from Reference 11, Table 11.5.2-1 and Table 11.5.5-1.
3-21 VER27
VEGP ODCM 3.3.2 Setpoint for the Final Noble Gas Monitor on Each Release Pathway 3.3.2.1 Overview of Method Gaseous effluent radioactivity monitors are intended to alarm prior to exceeding the limits of Section 3.1.2.a. Therefore, their alarm setpoints are established to ensure compliance with the following equation:
AG.SF. X*Rt c = the lesser of { (3.1)
AG*SF* X*Rk where:
c = the setpoint, in M-Ci/mL, of the radioactivity monitor measuring the concentration of radioactivity in the effluent line prior to release. The setpoint represents a concentration which, if exceeded, could result in dose rates exceeding the limits of Section 3.1.2.a at or beyond the SITE BOUNDARY.
AG = an administrative allocation factor applied to divide the release limit among all the gaseous release pathways at the site.
SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement.
x = the noble gas concentration for the release under consideration.
Rt = the ratio of the dose rate limit for the total body, 500 mrem/y, to the dose rate to the total body for the conditions of the release under consideration.
Rk = the ratio of the dose rate limit for the skin, 3000 mrem/y, to the dose rate to the skin for the conditions of the release under consideration.
Equation (3.1) shows the relationships of the critical parameters that determine the setpoint.
However, in order to apply the methodology presented in the equation to a mixture of noble gas radionuclides, radionuclide-specific concentrations and dose factors must be taken into account under conditions of maximum flowrate for the release point and annual average meteorology.
The basic setpoint method presented below is applicable to the radioactivity monitor nearest the pOint of release for the release pathway. For monitors measuring the radioactivity in source streams that merge with other streams prior to subsequent monitoring and release, the modifications presented in Section 3.3.3 must be applied.
3.3.2.2 Setpoint Calculation Steps Step 1: Determine the concentration, Xiv, of each noble gas radionuclide i in the gaseous stream v being considered for release, in accordance with the sampling and analysis requirements of Section 3.1.2. Then sum these concentrations to determine the total noble gas concentration, D<iV.
I 3-22 VER27
VEGPODCM Step 2: Determine Rb the ratio of the dose rate limit for the total body, 500 mrem/y, to the total body dose rate due to noble gases detected in the release under consideration, as follows:
Rt = 500 (3.2)
(X 10) ~JK;. av]
vb where:
500 = the dose rate limit for the total body, 500 mrem/y.
(X/O)vb = the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v. Table 3-4 includes an indication of what release elevation is applicable to each release pathway; release elevation determines the appropriate value of (X/O)Vb'
= the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mrem/y)/(JlCi/m 3 ), from Table 3-5.
= the release rate of noble gas radionuclide i from the release pathway under consideration. in JlCiis. calculated as the product of Xiv and fay. where:
= the concentration of noble gas radionuclide i for the particular release. in JlCi/mL.
= the maximum anticipated flowrate for release pathway v during the period of the release under consid~ration, in mUs.
Step 3: Determine Rk
the ratio of the dose rate limit for the skin, 3000 mremly, to the skin dose rate due to noble gases detected in the release under consideration, as follows:
Rk = 3000 (3.3)
( X/Q }vb 2: [(Li + 1.1Mi)* QiV]
where:
3000 = the dose rate limit for the skin, 3000 mrem/y.
~ = the skin dose factor due to beta emissions from noble gas radionuclide i, in (mrem/Y)/(JlCi/m3), from Table 3-5.
M = the air dose factor due to gamma emissions from noble gas radionuclide i, in (mradly)/(JlCi/m3). from Table 3-5.
1.1 = the factor to convert air dose in mrad to skin dose in mrem.
All other terms were defined previously.
3-23 VER27
VEGPODCM Step 4: Determine the maximum noble gas radioactivity monitor setpoint concentration.
Based on the values determined in previous steps, the radioactivity monitor setpoint for the planned release is calculated to ensure that the limits of Section 3.1.2.a will not be exceeded.
Because the radioactivity monitor responds primarily to radiation from noble gas radionuclides, the monitor setpoint Cnv (in IlCi/ml) is based on the concentration of all noble gases in the waste stream, as follows:
where:
cnv = the calculated setpoint, in IlCi/ml, for the noble gas monitor serving gaseous release pathway v.
AGv *SF,"[.Xiv *Rt e nv = the lesser of{ .. i . (3.4)
AG v SF "[. XiV Rk i
AG v = the administrative allocation factor for gaseous release pathway v, applied to divide the release limit among all the gaseous release pathways at the site.
The allocation factor may be assigned any value between 0 and 1, under the condition that the sum of the allocation factors for all Simultaneously-active final release pathways at the entire plant site does not exceed 1. Alternative methods for determination of AG v are presented in Section 3.3.4.
SF = the safety factor selected to compensate for statistical fluctuations and errors of measurement. The value for the safety factor must be between 0 and 1.
A value of 0.5 is reasonable for gaseous releases; a more precise value may be developed if desired.
Xiv = the measured concentration of noble gas radionuclide i in gaseous stream v, as defined in Step 1, in ,..,Ci/mL The values of RI and Rk to be used in the calculation are those which were determined in Steps 2 and 3 above.
Step 5: Determine whether the release is permissible, as follows:
If cnv :;:: D<iv, the release is permissible. However, if Cny is within about 10 percent of D<iv. it I j may be impractical to use this value of cnv
This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to adjust the allocation of the limits among the active release points. The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions.
If Cnv < D<iv. the release may not be made as planned. Consider the alternatives I
discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.
3-24 VER27
VEGPODCM 3.3.2.3 Use of the Calculated Setpoint The setpoint calculated above is in the units J.lCi/ml. The monitor actually measures a count rate, subtracts a predetermined background count rate, and multiplies by a calibration factor to convert from count rate to J.lCilml.
Initial calibration by the manufacturer and Georgia Power Company of the gaseous effluent monitors specified in Section 3.1.1 utilized at least one NIST-traceable gaseous radionuclide source in the exact geometry of each production monitor. The point and gaseous sources used covered the beta particle end point energy range from 0.293 MeV to at least 1.488 MeV. The calibration factor is a function of the radionuclide mix in the gas to be released, and normally will be calculated for the monitor based on the results of the sample results from the laboratory gamma-ray spectrometer system. The mix-dependent calibration factor will be used as the gain factor in the PERMS monitor, or used to modify the calculated base monitor setpoint so that the default calibration factor in the PERMS monitor can be left unchanged.
Notwithstanding the initial calibration, monitor calibration data for conversion between count rate and concentration may include operational data obtained from determining the monitor response to stream concentrations measured by sample analysis.
In all cases, monitor background must be controlled so that the monitor is capable of responding to concentrations in the range of the setpoint value. Contributions to the monitor background may include any or all of the following factors: ambient background radiation, plant-related radiation levels at the monitor location (which may change between shutdown and power conditions), and internal background due to contamination of the monitor's sample chamber.
3.3.3 Setpoints for Noble Gas Monitors on Effluent Source Streams Table 3-4 lists certain gaseous release pathways as being source streams. As may be seen in the figures of Section 3.2, these are streams that merge with other streams, prior to passing a final radioactivity monitor and being released. Unlike the final monitors, the source stream monitors measure radioactivity in effluent streams for which flow can be terminated; therefore, the source stream monitors have control logic to terminate the source stream release at the alarm setpoint.
3.3.3.1 Setpoint of the Monitor on the Source Stream Step 1: Determine the concentration Xis of each noble gas radionuclide i in source stream s (in J.lCilmL) according to the results of its required sample analyses [see Section 3.1.2].
Step 2: Determine rio the ratio of the dose rate limit for the total body, 500 mrem/y, to the total body dose rate due to noble gases detected in the source stream under consideration. Use the Xis values and the maximum antiCipated source stream flowrate fas in equation (3.2) to determine the total body dose rate for the source stream, substituting rt for Rt.
The SITE BOUNDARY relative dispersion value used in Steps 2 and 3 for the source stream is the same as the (XjQ)Vb that applies to the respective merged stream. This is 3-25 VER27
VEGPODCM because the (X/Q) value is determined by the meteorology of the plant site and the physical attributes of the release pOint, and is unaffected by whether or not a given source stream is operating.
Step 3: Determine rk, the ratio of the dose rate limit for the skin, 3000 mrem/y, to the skin dose rate due to noble gases detected in the source stream under consideration.
Use the XiS values and the maximum anticipated source stream flow rate fas in equation (3.3) to determine the skin dose rate for the source stream, substituting rk for Rk.
Step 4: Determine the maximum noble gas radioactivity monitor setpoint concentration, as follows:
AGS *SF* L XiS' rt ens = the lesser of I (3.5)
{
AGs . SF* L Xis' rk i
where:
cns = the calculated setpoint (in llCi/mL) for the noble gas monitor serving gaseous source stream s.
AG s = the administrative allocation factor applied to gaseous source stream s. For a given final release point v, the sum of all the AG s values for source streams contributing to the final release point must not exceed the release point's allocation factor Agv*
XiS = the measured concentration of noble gas radionuclide i in gaseous source stream s, as defined in Step 1, in llCi/mL.
The values of rl and rk to be used in the calculation are those which were determined in Steps 2 and 3 above. The safety factor, SF, was defined previously.
Step 5: Determine whether the release is permissible, as follows:
If C ns .2:. D<iS, the release is permissible. However, if cns is within about 10 percent of D<iS, it I I may be impractical to use this value of cns
This situation indicates that measured concentrations are approaching values which would cause the limits of Section 3.1.2.a to be exceeded. Therefore, steps should be taken to reduce contributing source terms of gaseous radioactive material, or to adjust the allocation of the limits among the active release pOints. The setpoint calculations (steps 1-4) must then be repeated with parameters that reflect the modified conditions.
If Cns < D<iS, the release may not be made as planned. Consider the alternatives I
discussed in the paragraph above, and calculate a new setpoint based on the results of the actions taken.
3-26 VER27
VEGPODCM 3.3.3.2 Effect on the Setpoint of the Monitor on the Merged Stream Before beginning a release from a monitored source stream, a setpoint must be determined for the source stream monitor as presented in Section 3.3.3.1. In addition, whether or not the source stream has its own effluent monitor, the previously-determined maximum allowable setpoint for the downstream final monitor on the merged stream must be redetermined. This is accomplished by repeating the steps of Section 3.3.2, with the following modifications.
Modification 1: The new maximum anticipated flowrate of the merged stream is the sum of the old merged stream maximum flowrate, and the maximum flowrate of the source stream being considered for release.
(3.6)
Modification 2: The new concentration of noble gas radionuclide i in the merged stream includes both the contribution of the merged stream without the source stream, and the source stream being considered for release.
(X.) = (fav)Old . (XiV )Old + fas + Xis IV new (fav )new (3.7) 3.3.4 Determination of Allocation Factors, AG When simultaneous gaseous releases are conducted, an administrative allocation factor must be applied to divide the release limit among the active gaseous release pathways. This is to assure that the dose rate limit for areas at and beyond the SITE BOUNDARY (see Section 3.1.2) will not be exceeded by simultaneous releases. The allocation factor for any pathway may be assigned any value between 0 and 1, under the following two conditions:
1. The sum of the allocation factors for all simultaneously-active final release paths at the plant site may not exceed 1.
2. The sum of the allocation factors for all simultaneously-active source streams merging into a given final release pathway may not exceed the allocation factor of that final release pathway.
Any of the following three methods may be used to assign the allocation factors to the active gaseous release pathways:
1. For ease of implementation, AGv may be equal for all release pathways:
1 AG ::- (3.8)
V N where:
N = the number of simultaneously active gaseous release pathways.
2. AG v for a given release pathway may be selected based on an estimate of the portion of the total SITE BOUNDARY dose rate (from all simultaneous releases) that is contributed 3-27 VER27
VEGP ODCM by the release pathway. During periods when a given building or release pathway is not subject to gaseous radioactive releases, it may be assigned an allocation factor of zero.
3. AG v for a given release pathway may be selected based on a calculation of the portion of the total SITE BOUNDARY dose rate that is contributed by the release pathway, as follows:
(3.9) where:
(XjO)Vb = the annual average SITE BOUNDARY relative concentration applicable to the gaseous release pathway v for which the allocation factor is being determined, in s/m 3*
~ = the total-body dose factor due to gamma emissions from noble gas radionuclide i, in (mrem/y)/(jlCi/m 3 ), from Table 3-5.
QjV = the release rate of noble gas radionuclide i from release pathway v, in jlCi/s, calculated as the product of Xiv and faY, where:
Xiv = the concentration of noble gas radionuclide i applicable to the gaseous release pathway v for which the allocation factor is being determined, in jlCilmL fay = the discharge flowrate applicable to gaseous release pathway v for which the allocation factor is being determined, in mUs.
(XjO)ro = the annual average SITE BOUNDARY relative concentration applicable to active gaseous release pathway r, in s/m 3
Qir = the release rate of noble gas radionuclide i applicable to active release pathway r, in jlCi/s, calculated as the product of Xir and far, where:
Xir = the concentration of noble gas radionuclide i applicable to active gaseous release pathway r, in jlCi/mL far = the discharge flowrate applicable to active gaseous release pathway r, in mUs.
N = the number of simultaneously active gaseous release pathways (including pathway v that is of interest).
NOTE: Although equations (3.8) and (3.9) are written to illustrate the assignment of the allocation factors for final release pathways, they may also be used to assign allocation factors to the source streams that merge into a given final release pathway.
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VEGPODCM 3.3.5 Setpoints for Noble Gas Monitors with Special Requirements At present, VEGP has no noble gas monitors for which setpoint methodologies are to be presented in the ODCM, and that require methods other than those in Section 3.3.2 or Section 3.3.3.
3.3.6 Setpoints for Particulate and Iodine Monitors In accordance with Section 5.1.1 of NRC NUREG-0133 (Reference 1), the effluent controls of Section 3.1.1 do not require that the ODCM establish setpoint calculation methods for particulate and iodine monitors. Therefore, the following is provided for information only: Initial setpoints for the particulate channels of effluent monitors RE-12442 and RE-2565 were determined as described in Reference 13.
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VEGPODCM 3.4 GASEOUS EFFLUENT COMPLIANCE CALCULATIONS 3.4.1 Dose Rates at and Beyond the Site Boundary Because the dose rate limits for areas at and beyond the SITE specified in Section 3.1.2 are site limits applicable at any instant in time, the summations extend over all simultaneously active gaseous final release pathways at the plant site. Table 3-4 identifies the gaseous final release pathways at the plant site, and indicates the (X/Otb value for each.
3.4.1.1 Dose Rates Due to Noble Gases For the purpose of implementing the controls of Section 3.1.2.a, the dose rates due to noble gas radionuclides in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:
For total body dose rates:
ORt = ~{(x /O)Vb 7[Ki O;vl} (3.10)
For skin dose rates:
(3.11) where:
DR t = the total body dose rate at the time of the release, in mrem/y.
DRk ::: the skin dose rate at the time of the release, in mrem/y.
Q iv ::: the release rate of noble gas radionuclide i, in jlCi/s, equal to the product of ftv and Xiv, where:
ftv = the actual average flowrate for release pathway v during the period of the release, in mUs.
All other terms were defined previously.
3.4.1.2 Dose Rates Due to lodine-131, lodine-133, Tritium, and Radionuclides in Particulate Form with Half-Lives Greater than 8 Days For the purpose of implementing the controls of Section 3.1.2.b, the dose rates due to lodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in areas at or beyond the SITE BOUNDARY, due to releases of gaseous effluents, shall be calculated as follows:
(3.12) where:
ORo = the dose rate to organ 0 at the time of the release, in mrem/y.
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VEGPODCM Pjo = the site-specific dose factor for radionuclide i and organ 0, in (mrem/y)/(IlCVm3). Since the dose rate limits specified in Section 3.1.2.b apply only to the child age group exposed to the inhalation pathway, the values of Pio may be obtained from Table 3-9, "Raipj for Inhalation Pathway, Child Age Group."
Q'iv = the release rate of radionuclide i from gaseous release pathway v, in IlCi/s.
For the purpose of implementing the controls of Section 3.1.2.b, only 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation.
All other terms were defined previously.
3.4.2 Noble Gas Air Dose at or Beyond Site Boundary For the purpose of implementing the controls of Section 3.1.3, air doses in areas at or beyond the SITE BOUNDARY due to releases of noble gases from each unit shall be calculated as follows (adapted from Reference 1, page 28, by including only long-term releases):
Dp =3.17XlO-8~{(X/Q)vb ~[Ni <QJ} (3.13)
Dr = 3.17X10-8~{(X IQ)vb t[M i < QiV ~ (3.14) where:
3.17 x 10-<1 = a units conversion factor: 1 y/(3.15 x 107 s).
D~ = the air dose due to beta emissions from noble gas radionuclides, in mrad.
Dy = the air dose due to gamma emissions from noble gas radionuclides, in mrad.
Ni = the air dose factor due to beta emissions from noble gas radionuclide i (mrad/y)/(IlCi/m3), from Table 3-5.
M = the air dose factor due to gamma emissions from noble gas radionuclide i (mradly)/(IlCi/m3), from Table 3-5.
QiV = the cumulative release of noble gas radionuclide i from release pathway v (IlCi), during the period of interest.
and all other terms are as defined above.
Because the air dose limit is on a per-reactor-unit basis, the summations extend over aI/ gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned to the two 3-31 VER27
VEGPODCM units in any reasonable manner, provided that all activity released via the particular shared release pathway is apportioned to one or the other unit.
The gaseous final release pathways at the plant site, and the (X/Q)Vb for each, are identified in Table 3-4.
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VEGPODCM Table 3-5. Dose Factors for Exposure to a Semi-Infinite Cloud of Noble Gases
'Y - Body (K) ~ - Skin (L) 'Y - Air (M) ~ - Air (N)
Nuclide (mrem/y) per (mrem/y) per (mradly) per (mrad/y) per (IlCi/m3) (IlCi/m3) (IlCi/m3) (IlCi/m3)
Kr-83m 7.56 E-02 0.00 E+OO 1.93 E+01 2.88 E+02 Kr-85m 1.17 E+03 1.46 E+03 1.23 E+03 1.97 E+03 Kr-85 1.61 E+01 1.34 E+03 1.72 E+01 1.95 E+03 Kr-87 5.92 E+03 9.73 E+03 6.17 E+03 1.03 E+04 Kr-88 1.47 E+04 2.37 E+03 1.52 E+04 2.93 E+03 Kr-89 1.66 E+04 1.01 E+04 1.73 E+04 1.06 E+04 Kr-90 1.56 E+04 7.29 E+03 1.63 E+04 7.83 E+03 Xe-131m 9.15 E+01 4.76 E+02 1.56 E+02 1.11 E+03 Xe-133m 2.51 E+02 9.94 E+02 3.27 E+02 1.48 E+03 Xe-133 2.94 E+02 3.06 E+02 3.53 E+02 1.05 E+03 Xe-135m 3.12 E+03 7.11 E+02 3.36 E+03 7.39 E+02 Xe-135 1.81 E+03 1.86 E+03 1.92 E+03 2.46 E+03 Xe-137 1.42 E+03 1.22 E+04 1.51 E+03 1.27 E+04 Xe-138 8.83 E+03 4.13 E+03 9.21 E+03 4.75 E+03 Ar-41 8.84 E+03 2.69 E+03 9.30 E+03 3.28 E+03 All values in this table were obtained from Reference 3 (Table B-1), with units converted.
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VEGPODCM Table 3-6. Dose Factors for Exposure to Direct Radiation from Noble Gases in an Elevated Finite Plume The contents of this table are not applicable to VEGP.
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VEGP ODCM 3.4.3 Dose to a Member of the Public at or Beyond Site Boundary The dose received by an individual due to gaseous releases from each reactor unit, to areas at or beyond the SITE BOUNDARY, depends on the individual's location, age group, and exposure pathways. The MEMBER OF THE PUBLIC expected to receive the highest dose in the plant vicinity is referred to as the contrOlling receptor. The dosimetrically-significant attributes of the currently-defined controlling receptor are presented in Table 3-7.
Doses to a MEMBER OF THE PUBLIC due to gaseous releases of 1-131,1-133, tritium, and all radionuclides in particulate form from each unit shall be calculated as follows (equation adapted from Reference 1, page 29, by considering only long-term releases):
(3.15) where:
Dja = the dose to organ j of an individual in age group a, due to gaseous releases of 1-131,1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days, in mrem.
3.17 x 10-8 = a units conversion factor: 1 y/(3.15 x 107 s).
Raipj = the site-specific dose factor for age group a, radionuclide i, exposure pathway p, and organ j. For the purpose of implementing the controls of Section 3.1.4, the exposure pathways applicable to calculating the dose to the currently-defined contrOlling receptor are included in Table 3-7; values of Raip] for each exposure pathway and radionuclide applicable to calculations of dose to the controlling receptor are included in Tables 3-8 through 3-12.
A detailed discussion of the methods and parameters used for calculating Raipj for the plant site is presented in Chapter 9. That information may be used for recalculating the RaiP] values if the underlying parameters change, or for calculating Raipj values for special radionuclides and age groups when performing the assessments discussed in Section 3.4.4 below.
WViP = the annual average relative dispersion or deposition at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and radionuclide i.
For all tritium pathways, and for the inhalation of any radionuclide: WViP is (x/otp ,the annual average relative dispersion factor for release pathway v, at the location of the contrOlling receptor (slm 3). For the ground-plane exposure pathway, and for all ingestion-related pathways for radionuclides other than tritium: Wvip is (D/O)..p ,the annual average relative deposition factor for release pathway v, at the location of the contrOlling receptor (m-2).
Values of (X/O)vp and (D/O)vp for use in calculating the dose to the currently-defined contrOlling receptor are included in Table 3-7.
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VEGPODCM Q'iV = the cumulative release of radionuclide i from release pathway v, during the period of interest (IlCi). For the purpose of implementing the controls of Section 3.1.4, only 1-131,1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days should be included in this calculation.
In any dose assessment using the methods of this subsection, only radionuclides detectable above background in their respective samples should be included in the calculation.
Because the member of the public dose limit is on a per-unit basis, the summations extend over all gaseous final release pathways for a given unit. For a release pathway discharging materials originating in both reactor units, the activity discharged from the release point may be apportioned between the two units in any reasonable manner, provided that all activity released from the plant site is apportioned to one or the other unit.
The gaseous final release pathways at the plant site, and the release elevation for each, are identified in Table 3-4.
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VEGPODCM Table 3-7. Attributes of the Controlling Receptor The locations of members of the public in the vicinity of the plant site, and the exposure pathways associated with those locations, are determined in the Annual Land Use Census.
Dispersion and deposition values were calculated based on site meteorological data collected for the period January 1, 1985 through December 31, 1987.
Based on an analysis of this information, the current controlling receptor for the plant site is described as follows.
Sector: WSW Distance: 1.2 miles Age Group: Child Exposure Pathways: Inhalation, ground plane, cow meat, and garden vegetation Dispersion Factors (X/OJvb:
Ground-Level release points: 6.20 E-7 s/m 3 Mixed-Mode release points: 1.27 E-7 s/m 3 Deposition Factors (D/O)vb:
Ground-Level release points: 2.80 E-9 m-2 Mixed-Mode release pOints: 9.90 E-10 m-2 3-37 VER27
VEGPOOCM 3.4.4 Dose Calculations to Support Other Requirements Case 1: A radiological impact assessment may be required to support evaluation of a reportable event.
Dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the dispersion and deposition parameters [(X/O) and (0/0)] for the period covered by the report, and using the appropriate pathway dose factors (Raipj ) for the receptor of interest. Methods for calculating (X/O) and (0/0) from meteorological data are presented in Chapter 8.
Values of Raipj other than those presented in Tables 3*8 through 3*12 may need to be calculated. Methods and parameters for calculating values of Raipj are presented in Chapter 9. When calculating Raipj for evaluation of an event, pathway and usage factors specific to the receptor involved in the event may be used in place of the values in Chapter 9, if the specific values are known.
Case 2: A dose calculation is required to evaluate the results of the Land Use Census, under the provisions of Section 4.1.2.
In the event that the Land Use Census reveals that exposure pathways have changed at previously* identified locations. or if new locations are identified, it may be necessary to calculate doses at two or more locations to determine which should be designated as the controlling receptor. Such dose calculations may be performed using the equations in Section 3.4.3, with the substitution of the annual average dispersion and deposition values
[(X/O) and (0/0)] for the locations of interest, and using the appropriate pathway dose factors (Raipj) for the receptors of interest.
Methods for calculating (X/O) and (0/0) from meteorological data are presented in Chapter 8. The values of Raipj other than those presented in Tables 3*8 through 3*12 may need to be calculated. Methods and parameters for calculating values of RaiPi are presented in Chapter 9.
Case 3: Under Section 5.2, a dose calculation may be required to support the determination of a component of the total dose to a receptor other than that currently defined as the controlling receptor.
Dose calculations would be performed using the equations in Section 3.4.3, with the dispersion and deposition parameters and appropriate values of (RaiPj ) for the receptor of interest.
Appropriate values of the dispersion and deposition parameters, if not found in Table 3-7, would need to be calculated. Methods for calculating (X/O) and (0/0) from meterological data are presented in Chapter 8.
Appropriate values of Raipj
if not found in Tables 3-8 through 3-12, would need to be calculated. Methods and parameters for calculating values of RaiPj are presented in Chapter 9.
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VEGPODCM Table 3-S. Raipj for Ground Plane Pathway, All Age Groups Nuclide T. Body Skin H-3 0.00 0.00 C-14 0.00 0.00 P-32 0.00 0.00 Cr-51 4.66E+06 5.51E+06 Mn-54 1.39E+09 1.63E+09 Fe-55 0.00 0.00 Fe-59 2.73E+OS 3.21E+OS Co-5S 3.79E+OS 4.44E+OS Co-60 2.15E+10 2.53E+10 Ni-63 0.00 0.00 Zn-65 7.47E+OS S.59E+OS Rb-S6 S.99E+06 1.03E+07 Sr-S9 2.16E+04 2.51E+04 Sr-90 0.00 0.00 Y-91 1.07E+06 1.21E+06 Zr-95 2.45E+OS 2.S4E+OS Nb-95 1.37E+OS 1.61E+OS Ru-103 1.0SE+OS 1.26E+OS Ru-106 4.22E+OS 5.07E+OS Ag-110m 3.44E+09 4.01E+09 Sb-124 5.9SE+OS 6.90E+OS Sb-125 2.34E+09 2.64E+09 Te-125m 1.55E+06 2.13E+06 Te-127m 9.16E+04 1.0SE+05 Te-129m 1.9SE+07 2.31E+07 1-131 1.72E+07 2.09E+07 1-133 2.45E+06 2.9SE+06 Cs-134 6.S6E+09 S.00E+09 Cs-136 1.51 E+OS 1.71E+08 Cs-137 1.03E+10 1.20E+10 Ba-140 2.05E+07 2.35E+07 Ce-141 1.37E+07 1.54E+07 Ce-144 6.95E+07 S.04E+07 Pr-143 0.00 0.00 Nd-147 S.39E+06 1.01E+07
1. Units are m2*(mrem/yr)/(IlCi/s).
2. The values in the Total Body column also apply to the Bone, Liver, Thyroid, Kidney, Lung, and GI-LLI organs.
3. This table also supports the calculations of section 6.2.
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VEGPODCM Table 3-9. Raipj for Inhalation Pathway, Child Age Group ucli IT~ T Y Lung GI-LLI H-3 0.00 1.12E+03 1.12E+03 1.12E+03 1. 12E+03 1.12E+03 1.12E+03 C-14 3.59E+04 S.73E+03 S.73E+03 S.73E+03 S.73E+03 S.73E+03 S.73E+03 P-32 2.S0E+OS 1.14E+05 9.88E+04 0.00 0.00 0.00 4.22E+04 Cr-51 0.00 0.00 1.54E+02 8.55E+01 2.43E+01 1.70E+04 1.08E+03 Mn-54 0.00 4.29E+04 9.51 E+03 0.00 1.00E+04 1.58E+OS 2.29E+04 Fe-55 4.74 E+04 2.52E+04 7.77E+03 0.00 0.00 1.11E+05 2.87E+03 Fe-59 2.07E+04 3.34E+04 1.S7E+04 0.00 0.00 1.27E+OS 7.07E+04 Co-58 0.00 1.77E+03 3.1SE+03 0.00 0.00 1.11E+OS 3.44E+04 Co-SO 0.00 1.31E+04 2.2SE+04 0.00 0.00 7.07E+OS 9.S2E+04 Ni-S3 8.21E+05 4.S3E+04 2.80E+04 0.00 0.00 2.75E+05 S.33E+03 Zn-S5 4.2SE+04 1.13E+05 7.03E+04 0.00 7.14E+04 9.95E+05 1.S3E+04 Rb-8S 0.00 1.98E+05 1.14E+05 0.00 0.00 0.00 7.99E+03 Sr-89 5.99E+05 0.00 1.72E+04 0.00 0.00 2.1SE+OS 1.S7E+05 Sr-90 1.01E+08 0.00 S.44E+OS 0.00 0.00 1.48E+07 3.43E+05 Y-91 9.14E+05 0.00 2.44E+04 0.00 0.00 2.S3E+OS 1.84E+05 Zr-95 1.90E+05 4.18E+04 3.70E+04 0.00 5.9SE+04 2.23E+OS S.11 E+04 Nb-95 2.35E+04 9.18E+03 S.55E+03 0.00 8.S2E+03 S.14E+05 3.70E+04 Ru-103 2.79E+03 0.00 1.07E+03 0.00 7.03E+03 S.S2E+05 4.48E+04 Ru-10S 1.3SE+05 0.00 1.S9E+04 0.00 1.84E+05 1.43E+07 4.29E+05 Ag-110m 1.S9E+04 1.14E+04 9.14E+03 0.00 2.12E+04 5.48E+OS 1.00E+05 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m S.73E+03 2.33E+03 9.14E+02 1.92E+03 0.00 4.77E+05 3.38E+04 Te-127m 2.49E+04 8.55E+03 3.02 E+03 S.07E+03 S.3SE+04 1.48E+OS 7.14E+04 Te-129m 1.92E+04 S.85E+03 3.04E+03 S.33E+03 5.03E+04 1.7SE+OS 1.82E+05 1-131 4.81E+04 4.81E+04 2.73E+04 1.S2E+07 7.88E+04 0.00 2.84E+03 1-133 1.SSE+04 2.03E+04 7.70E+03 3.85E+OS 3.38E+04 0.00 5.48E+03 Cs-134 S.51 E+05 1.01E+OS 2.25E+05 0.00 3.30E+05 1.21E+05 3.85E+03 Cs-13S S.51E+04 1.71 E+05 1.1SE+05 0.00 9.SSE+04 1.4SE+04 4.18E+03 Cs-137 9.07E+05 8.2SE+05 1.28E+05 0.00 2.82E+05 1.04E+05 3.S2E+03 8a-140 7.40E+04 6.48E+01 4.33E+03 0.00 2.11 E+01 1.74E+OS 1.02E+OS Ce-141 3.92E+04 1.95E+04 2.90E+03 0.00 8.55E+03 5.44E+05 5.SSE+04 Ce-144 S.77E+06 2.12E+06 3.61E+OS 0.00 1.17E+OS 1.20E+07 3.89E+05 Pr-143 1.85E+04 S.55E+03 9.14E+02 0.00 3.00E+03 4.33E+05 9.73E+04 d-147 1.08E+04 8.73E+03 S.81E+02 0.00 4.81E+03 3.28E+05 8.21E+04
1. Units are (mrem/yr}/(J,lCi/m 3 ) for all radionuclides.
2. This table also supports the calculations of section S.2.
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VEGP ODCM Table 3-10. Raipj for Inhalation Pathway, Adult Age Group Bone Liver T.Body Thyroid Kidney Lung GI-LLI H-3 0.00 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 1.26E+03 C-14 1.82E+04 3,41 E+03 3,41E+03 3,41E+03 3,41E+03 3,41E+03 3,41 E+03 P-32 1.32E+06 7.71E+04 5.01E+04 0.00 0.00 0.00 8.64E+04 Cr-51 0.00 0.00 1.00E+02 5.95E+01 2.28E+01 1,44E+04 3.32E+03 Mn-54 0.00 3.96E+04 6.30E+03 0.00 9.84E+03 1,40E+06 7. 74E+04 Fe-55 2,46E+04 1.70E+04 3.94E+03 0.00 0.00 7.21E+04 6.03E+03 Fe-59 1.18E+04 2.78E+04 1.06E+04 0.00 0.00 1.02E+06 1.88E+05 Co-58 0.00 1.58E+03 2.07E+03 0.00 0.00 9.28E+05 1.06E+05 Co-60 0.00 1.15E+04 1,48E+04 0.00 0.00 5.97E+06 2.85E+05 Ni-63 4.32E+05 3.14E+04 1,45E+04 0.00 0.00 1.78E+05 1.34E+04 Zn-65 3.24E+04 1.03E+05 4.66E+04 0.00 6.90E+04 8.64E+05 5.34E+04 Rb-86 0.00 1.35E+05 5.90E+04 0.00 0.00 0.00 1.66E+04 Sr-89 3.04E+05 0.00 8.72E+03 0.00 0.00 1,40E+06 3.50E+05 Sr-90 9.92E+07 0.00 6.10E+06 0.00 0.00 9.60 E+06 7.22E+05 Y-91 4.62E+05 0.00 1.24E+04 0.00 0.00 1.70E+06 3.85E+05 Zr-95 1.07E+05 3.44E+04 2.33E+04 0.00 5,42E+04 1.77E+06 1.50E+05 Nb-95 1,41 E+04 7.82 E+03 4.21E+03 0.00 7.74E+03 5.05E+05 1.04E+05 Ru-103 1.53E+03 0.00 6.58E+02 0.00 5.83E+03 5.05E+05 1.10E+05 Ru-106 6.91 E+04 0.00 8.72E+03 0.00 1.34E+05 9.36E+06 9.12E+05 Ag-110m 1.08E+04 1.00E+04 5.94E+03 0.00 1.97E+04 4.63E+06 3.02E+05 Sb-124 3.12E+04 5.89E+02 1.24E+04 7.55E+01 0.00 2,48E+06 4.06E+05 Sb-125 6.61E+04 7.13E+02 1.33E+04 5.87E+01 0.00 2.20E+06 1.01 E+05 Te-125m 3,42E+03 1.58E+03 4.67E+02 1.05E+03 1.24E+04 3.14E+05 7.06E+04 Te-127m 1.26E+04 5.77E+03 1.57E+03 3.29E+03 4.58E+04 9.60E+05 1.50E+05 Te-129m 9.76E+03 4.67E+03 1.58E+03 3,44E+03 3.66E+04 1.16E+06 3.83E+05 1-131 2.52E+04 3.58E+04 2.05E+04 1.19E+07 6.13E+04 0.00 6.28E+03 1-133 8.64E+03 1,48E+04 4.52E+03 2.15E+06 2.58E+04 0.00 8.88E+03 Cs-134 3.73E+05 8,48E+05 7.28E+05 0.00 2.87E+05 9.76E+04 1.04E+04 Cs-136 3.90E+04 1,46E+05 1.10E+05 0.00 8.56E+04 1.20E+04 1.17E+04 Cs-137 4.78E+05 6.21 E+05 4.28E+05 0.00 2.22E+05 7.52E+04 8,40E+03 Ba-140 3.90E+04 4.90E+01 2.57E+03 0.00 1.67E+01 1.27E+06 2.18E+05 Ce-141 1.99E+04 1.35E+04 1.53E+03 0.00 6.26E+03 3.62E+05 1.20E+05 Ce-144 3,43E+06 1,43E+06 1.84E+05 0.00 8,48E+05 7.78E+06 8.16E+05 Pr-143 9.36E+03 3.75E+03 4.64E+02 0.00 2.16E+03 2.81E+05 2.00E+05 Nd-147 5.27E+03 6.10E+03 3.65E+02 0.00 3.56E+03 2.21 E+05 1.73E+05
1. Units are (mrem/yr)/(J.lCi/m 3 ) for all radionuclides.
2. This table is included to support the calculations of section 6.2.
3-41 VER27
VEGP ODCM Table 3-11. Raipj for Cow Meat Pathway, Child Age Group Liver T. Body Thyroid Kidney Lung GI-LLI H-3 0.00 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 2.34E+02 C-14 S.29E+OS 1.06E+OS 1.06E+OS 1.06E+OS 1.06E+OS 1.06E+OS 1.06E+OS P-32 7.41E+09 3.47E+OS 2.S6E+OS 0.00 0.00 0.00 2.0SE+OS Cr-S1 0.00 0.00 S.79E+03 4.SSE+03 1.33E+03 S.91E+03 4.66E+OS Mn-S4 0.00 S.01E+06 2.13E+06 0.00 2.2SE+06 0.00 6.72E+06 Fe-SS 4.S7E+OS 2.42E+OS 7.S1E+07 0.00 0.00 1.37E+OS 4.49E+07 Fe-S9 3.76E+OS 6.09E+OS 3.03E+OS 0.00 0.00 1.77E+OS 6.34E+OS Co-SS 0.00 1.64E+07 S.02E+07 0.00 0.00 0.00 9.SSE+07 Co-60 0.00 6.93E+07 2.04E+OS 0.00 0.00 0.00 3.S4E+OS Ni-63 2.91E+10 1.S6E+09 9.91E+OS 0.00 0.00 0.00 1.0SE+OS Zn-6S 3.7SE+OS 1.00E+09 6.22E+OS 0.00 6.30E+OS 0.00 1.76E+OS Rb-S6 0.00 S.77E+OS 3.SSE+OS 0.00 0.00 0.00 3.71E+07 Sr-S9 4.S2E+OS 0.00 1.3SE+07 0.00 0.00 0.00 1.S7E+07 Sr-90 1.04E+10 0.00 2.64E+09 0.00 0.00 0.00 1.40E+OS Y-9l 1.S0E+06 0.00 4.S2E+04 0.00 0.00 0.00 2.40E+OS Zr-9S 2.66E+06 S.SSE+OS S.21E+OS 0.00 S.3SE+OS 0.00 6.11E+OS Nb-9S 3.l0E+06 1.21E+06 S.62E+OS 0.00 1.13E+06 0.00 2.23E+09 Ru-l03 1.SSE+OS 0.00 S.96E+07 0.00 3.90E+OS 0.00 4.01E+09 Ru-l06 4.44E+09 0.00 S.S4E+OS 0.00 S.99E+09 0.00 6.90E+l0 Ag-110m S.39E+06 S.67E+06 4.S3E+06 0.00 1.06E+07 0.00 6.74E+OS Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-12S 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-12Sm S.69E+OS 1.S4E+OS 7.S9E+07 1.60E+OS 0.00 0.00 S.49E+OS Te-127m 1.77E+09 4.7SE+OS 2.11E+OS 4.24E+OS S.06E+09 0.00 1.44E+09 Te-129m 1.79E+09 S.OOE+OS 2.7SE+OS S.77E+OS S.26E+09 0.00 2.1SE+09 1-131 1.6SE+07 1.66E+07 9.46E+06 S.SOE+09 2.73E+07 0.00 1.4SE+06 1-133 S.67E-01 7.02E-01 2.66E-01 1.30E+02 1.17E+00 0.00 2.S3E-01 Cs-134 9.22E+OS 1.S1 E+09 3.19E+OS 0.00 4.69E+OS 1.6SE+OS S.16E+06 Cs-136 1.62E+07 4.46E+07 2.88E+07 0.00 2.37E+07 3.S4E+06 1.S7E+06 Cs-137 1.33E+09 1.28E+09 1.SSE+OS 0.00 4.16E+OS 1.S0E+OS 7.99E+06 8a-140 4.3SE+07 3.S4E+04 2.S6E+06 0.00 1.2SE+04 2.29E+04 2.22E+07 Ce-141 2.22E+04 1.11E+04 1.64E+03 0.00 4.S6E+03 0.00 1.38E+07 Ce-144 2.32E+06 7.26E+OS 1.24E+OS 0.00 4.02E+OS 0.00 1.89E+OS Pr-143 3.34 E+04 1.00E+04 1.66E+03 0.00 S.43E+03 0.00 3.60E+07 Nd-147 1.17E+04 9.47E+03 7.33E+02 0.00 S.19E+03 0.00 1.S0E+07 Units are (mrem/yr)/(J,lCi/m 3) for tritium, and m2 .(mrem/yr)/(J,lCi/s) for all other radionuclides.
3-42 VER27
VEGPODCM Table 3-12. Raipj for Garden Vegetation Pathway, Child Age Group
~ide 80ne Liver T.80dy Thyroid Kidney Lung GI-LLI H-3 0.00 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 4.01E+03 C-14 S.S9E+OS 1.7SE+OS 1.7SE+OS 1.7SE+OS 1.7SE+OS 1.7SE+OS 1.7SE+OS P-32 3.37E+09 1.5SE+OS 1.30E+OS 0.00 0.00 0.00 9.31E+07 Cr-51 0.00 0.00 1.17E+05 6.50E+04 1.7SE+04 1.19E+05 6.21E+06 Mn-54 0.00 6.65E+OS 1.77E+OS 0.00 1.S6E+OS 0.00 5.5SE+OS Fe-55 S.01E+OS 4.25E+OS 1.32E+OS 0.00 0.00 2.40E+OS 7.S7E+07 Fe-59 3.9SE+OS 6.43E+OS 3.20E+OS 0.00 0.00 1.S6E+OS 6.70E+OS Co-5S 0.00 6.44E+07 1.97E+OS 0.00 0.00 0.00 3.76E+OS Co-60 0.00 3.7SE+OS 1.12E+09 0.00 0.00 0.00 2.10E+09 Ni-63 3.95E+10 2.11E+09 1.34E+09 0.00 0.00 0.00 1.42E+OS Zn-65 S.13E+OS 2.16E+09 1.35E+09 0.00 1.36E+09 0.00 3.S0E+OS Rb-S6 0.00 4.52E+OS 2.7SE+OS 0.00 0.00 0.00 2.91E+07 Sr-S9 3.60E+10 0.00 1.03E+09 0.00 0.00 0.00 1.39E+09 Sr-90 1.24E+12 0.00 3.15E+11 0.00 0.00 0.00 1.67E+10 Y-91 1.S6E+07 0.00 4.99E+05 0.00 0.00 0.00 2.4SE+09 Zr-95 3.S6E+06 S.4SE+05 7.55E+05 0.00 1.21E+06 0.00 B.S5E+OS Nb-95 4.10E+05 1.60E+05 1.14E+05 0.00 1.50E+05 0.00 2.96E+OS Ru-103 1.53E+07 0.00 5.90E+06 0.00 3.S6E+07 0.00 3.97E+OS Ru-106 7.45E+OS 0.00 9.30E+07 0.00 1.01E+09 0.00 1.16E+10 Ag-110m 3.21E+07 2.17E+07 1.73E+07 0.00 4.04E+07 0.00 2.5SE+09 Sb-124 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Sb-125 0.00 0.00 0.00 0.00 0.00 0.00 0.00 Te-125m 3.51E+OS 9.50E+07 4.67E+07 9.S4E+07 0.00 0.00 3.3SE+OS Te-127m 1.32E+09 3.56E+OS 1.57E+OS 3.16E+OS 3.77E+09 0.00 1.07E+09 Te-129m S.41 E+OS 2.35E+OS 1.31E+OS 2.71 E+OS 2.47E+09 0.00 1.03E+09 1-131 1.43E+OS 1.44E+OS S.17E+07 4.75E+10 2.36E+OS 0.00 1.2BE+07 1-133 3.53E+06 4.37E+06 1.65E+06 S.11E+OS 7.2SE+06 0.00 1.76E+06 Cs-134 1.60E+10 2.63E+10 5.55E+09 0.00 S.15E+09 2.93E+09 1.42E+OS Cs-136 S.24E+07 2.27E+OS 1.47E+OS 0.00 1.21E+OS 1.S0E+07 7.96E+06 Cs-137 2.39E+10 2.29E+10 3.3SE+09 0.00 7.46E+09 2.6SE+09 1.43E+OS 8a-140 2.77E+OS 2.42E+05 1.61E+07 0.00 7.S9E+04 1.45E+05 1.40E+OS Ce-141 6.56E+05 3.27E+05 4.S6E+04 0.00 1.43E+05 0.00 4.0SE+OS Ce-144 1.27E+OS 3.9SE+07 6.7SE+06 0.00 2.21E+07 0.00 1.04E+10 Pr-143 1.46E+05 4.37E+04 7.23E+03 0.00 2.37E+04 0.00 1.57E+OS Nd-147 7.15E+04 5.79E+04 4.4SE+03 0.00 3.1SE+04 0.00 9.17E+07 Units are (mrem/yr)/(IlCi/m3) for tritium, and m2'(mrem/yr)/(IlCi/s) for all other radionuclides.
3-43 VER27
VEGPODCM 3.5 GASEOUS EFFLUENT DOSE PROJECTIONS 3.5.1 Thirty-One Day Dose Projections In order to meet the requirements of the limit for operation of the gaseous radwaste treatment system (see Section 3.1.5), dose projections must be made at least once each 31 days; this applies during periods in which a discharge to areas at or beyond the SITE BOUNDARY of gaseous effluents containing radioactive materials occurs or is expected.
Projected 31-day air doses and doses to individuals due to gaseous effluents may be determined as follows:
For air doses:
DfJp =[ D; ]X31+D/h (3.16)
For individual doses:
Dap=.- I\ Doc) f t
x31+D aa (3.17) where:
D~p = the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases.
D!lc = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.
DIla = the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, DIla may be set to zero.
Dill = the projected air dose due to gamma emissions from noble gases for the next 31 days of gaseous releases.
D)'C = the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration.
D"tS = the anticipated air dose due to gamma emissions from noble gas releases, contributed by any planned activities during the next 31-day period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, D"tS may be set to zero.
3-44 VER27
VEGPODCM Dop = the projected dose to the total body or organ 0, due to releases of 1-131, 1-133, tritium, and particulates for the next 31 days of gaseous releases.
Doc = the cumulative dose to the total body or organ 0, due to releases of 1-131, 1-133, tritium, and particulates that have occurred in the elapsed portion of the current quarter, plus the release under consideration.
Dca = the anticipated dose to the total body or organ 0, due to releases of 1-131 ,
1-133, tritium, and particulates, contributed by any planned activities during the next 31~ay period, if those activities will result in gaseous releases that are in addition to routine gaseous effluents. If only routine gaseous effluents are anticipated, Doa may be set to zero.
t = the number of whole or partial days elapsed into the current quarter, including the time to the end of the release under consideration (even if the release continues into the next quarter).
3.5.2 Dose Projections for Specific Releases
\
Dose projections may be performed for a particular release by performing a pre-release dose calculation assuming that the planned release will proceed as anticipated. For air dose and individual dose projections due to gaseous effluent releases, follow the methodology of Section 3.4, using sample analysis results for the gaseous stream to be released, and parameter values expected to exist during the release period.
3-45 VER27
VEGP ODCM 3.6 DEFINITIONS OF GASEOUS EFFLUENT TERMS Section of Term Definition Initial Use AG= the administrative allocation factor for gaseous streams, applied to divide the gaseous release limit among all the release pathways [unitless]. 3.3.2.1 AG s = the administrative allocation factor for gaseous source stream s, applied to divide the gaseous release limit among all the release pathways (unitless]. 3.3.3 AG v = the administrative allocation factor for gaseous release pathway v, applied to divide the gaseous release limit among all the release pathways [unitless]. 3.3.2.2 c= the setpoint of the radioactivity monitor measuring the concentration of radioactivity in the effluent line prior to release
[IlCi/mL]. 3.3.2.1 Cns = the calculated noble gas effluent monitor setpoint for gaseous source stream s [IlCi/mL]. 3.3.3 Cnv = the calculated noble gas effluent monitor setpoint for release pathway v [IlCilmL]. 3.3.2.2 Dja = the dose to organ j of an individual in age group a, due to gaseous releases of 1-131,1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days (mrem]. 3.4.3 Doa = the anticipated dose to organ 0 due to releases of non-noble-gas radionuclides, contributed by any planned activities during the next 31-day period [mrem]. 3.5.1 the cumulative dose to organ 0 due to releases of non-noble-gas radionuclides that have occurred in the elapsed portion of the current quarter, plus the release under consideration (mrem]. 3.5.1 Dop= the projected dose to organ 0 due to the next 31 days of gaseous releases of non-noble-gas radionuclides (mrem]. 3.5.1 the air dose due to beta emissions from noble gas radionuclides (mrad]. 3.4.2 Dpa = the anticipated air dose due to beta emissions from noble gas releases, contributed by any planned activities during the next 31-day period [mrad]. 3.5.1 3-46 VER27
VEGPODCM Section of Term Definition Initial Use Dpc = the cumulative air dose due to beta emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrad]. 3.5.1 Dpp= the projected air dose due to beta emissions from noble gases, for the next 31 days of gaseous releases [mrad]. 3.5.1 Dr'" the air dose due to gamma emissions from noble gas radionuclides [mrad]. 3.4.2 D"<<I= the anticipated air dose due to gamma emissions from noble gas releases, contributed by any planned activities during the next 31-day period [mrad]. 3.5.1 D'jC = the cumulative air dose due to gamma emissions from noble gas releases that have occurred in the elapsed portion of the current quarter, plus the release under consideration [mrad]. 3.5.1 Dyp= the projected air dose due to gamma emissions from noble gases, for the next 31 days of gaseous releases [mrad]. 3.5.1 (D/Q)vp = the annual average relative deposition factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [m-2]. 3.4.3 DR k = the skin dose rate at the time of the release [mrem/y]. 3.4.1.1 DRo= the dose rate to organ 0 at the time of the release [mrem/y]. 3.4.1.2 DRt = the total body dose rate at the time of the release [mrem/y]. 3.4.1.1 fay = the maximum anticipated actual dischargeflowrate for release pathway v during the period of the planned release [mUs]. 3.3.2.2 fas = the maximum anticipated actual discharge flowrate for gaseous source stream s during the period of the planned release [mUs]. 3.3.3
~= the total body dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5 [(mrem/y)/()lCi/m 3)]. 3.3.2.2 Li= the skin dose factor due to beta emissions from noble gas radio-nuclide i, from Table 3-5 [(mrern/y)/()lCi/m 3 )]. 3.3.2.2 M= the air dose factor due to gamma emissions from noble gas radionuclide i, from Table 3-5 [(mrad/y)/()lCi/m 3)]. 3.4.2 3-47 VER27
VEGPODCM Section of Term Definition Initial Use N= the number of simultaneously active gaseolJs release pathways
[unitless). 3.3.4 Nj = the air dose factor due to beta emissions from noble gas radionuclide i, from Table 3-5 [(mradly)/()lCi/m 3)). 3.4.2 Plo = the site-specific dose factor for radionuclide i (1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) and organ o. The values of Pio are equal to the site-specific A aipj values presented in Table 3-9 [(mrem/y)/()lCilm3)). 3.4.1.2 OIV= the release rate of noble gas radionuclide i from release pathway v during the period of interest [)lCi/s). 3.3.2.2 O'lv= the release rate of radionuclide i (1-131, 1-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days) from gaseous release pathway v during the period of interest [)lCi/s). 3.4.1.2 OIV= the cumulative release of noble gas radionuclide i from release pathway v during the period of interest [)lCi). 3.4.2 O'IV= the cumulative release of non-nOble-gas radionuclide i from release pathway v, during the period of interest [)lCi). 3.4.3 Aaipj = the site-specific dose factor for age group a, radionuclide i, exposure pathway p, and organ j. Values and units of Aaipj for each exposure pathway, age group, and radionuclide that may arise in calculations for implementing Section 3.1.4 are listed in Table 3-8 through Table 3-9. 3.4.3 Ak = the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in the release under consideration
[unitless]. 3.3.2.1 At = the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the release under consideration [unitless). 3.3.2.1 rk = the ratio of the skin dose rate limit for noble gases, to the skin dose rate due to noble gases in the source stream under consideration [unitless). 3.3.3.1 rt = the ratio of the total body dose rate limit for noble gases, to the total body dose rate due to noble gases in the source stream under consideration [unitless). 3.3.3.1 3-48 VEA27
VEGPODCM Section of Term Definition Initial Use SF= the safety factor used in gaseous setpoint calculations to compensate for statistical fluctuations and errors of measurement
[unitless]. 3.3.2.2 t= the number of whole or partial days elapsed in the current quarter, including the period of the release under consideration. 3.5.1 Wvjp= the annual average relative dispersion [(X/otp]or deposition
[(D/O)vp] at the location of the controlling receptor, for release pathway v, as appropriate to exposure pathway p and 3.4.3 radionuclide i.
X= the noble gas concentration for the release under consideration
[J1,Ci/mL]. 3.3.2.1 Xir = the concentration of radionuclide i applicable to active gaseous release pathway r [J1,Ci/mL]. 3.3.4 Xis = the measured concentration of radionuclide i in gaseous source stream s [J1,Ci/mL]. 3.3.3 Xiv = the measured concentration of radionuclide i in gaseous stream v
[J1,CilmL). 3.3.2.2 (X/Q) = the highest relative concentration at any pOint at or beyond the SITE BOUNDARY [slm1. 3.3.2.1 (X/O)rb = the annual average SITE BOUNDARY relative concentration applicable to active gaseous release pathway r [s/m 3]. 3.3.4 (X/O)Vb = the highest annual average relative concentration at the SITE BOUNDARY for the discharge point of release pathway v, from Table 3-4 [slm 3]. 3.3.2.2 (X/O)vp = annual average relative dispersion factor for release pathway v, at the location of the controlling receptor, from Table 3-7 [51m 3 ]. 3.4.3 3-49 VER27
VEGPODCM CHAPTER 4 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 4.1 LIMITS OF OPERATION The following limits are the same for both units at the site. Thus, a single program including monitoring, land use survey, and quality assurance serves both units.
4.1.1 Radiological Environmental Monitoring The Radiological Environmental Monitoring Program (REMP) shall be conducted as specified in Table 4-1.
4.1.1.1 Applicability This control applies at all times.
4.1.1.2 Actions 4.1.1.2.1 With the REMP not being conducted as specified in Table 4-1 , submit to the Nuclear Regulatory Commission (NRC), in the Annual Radiological Environmental Operating Report, a description of the reasons for not conducting the program as required and the plans for preventing a recurrence. Deviations from the required sampling schedule are permitted if specimens are unobtainable due to hazardous conditions, unavailability, inclement weather, equipment malfunction, or other just reasons. If deviations are due to equipment malfunction, efforts shall be made to complete corrective action prior to the end of the next sampling period.
4.1.1.2.2 With the confirmed 1 measured level of radioactivity as a result of plant effluents in an environmental sampling medium specified in Table 4-1 exceeding the reporting levels of Table 4-2 when averaged over any calendar quarter, submit within 30 days a special report to the NRC.
The special report shall identify the cause(s) for exceeding the limit(s) and define the corrective action(s) to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Sections 2.1.3, 3.1.3, and 3.1 .4. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in the special report.
When more than one of the radionuclides in Table 4-2 are detected in the sampling medium, this report shall be submitted if:
concentration (l) concentration (2)
+ + ,,' ;:: 1.0 reporting level (1) reporting level (2)
Defined as confirmed by reanalysis of the original sample. or analysis of a duplicate or new sample, as appropriate. The results of the confirmatory analysis shall be completed at the earliest time consistent with the analysis.
4-1 VER27
VEGPODCM When radionuclides other than those in Table 4-2 are detected and are the result of plant effluents, this special report shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits stated in Sections 2.1.3, 3.1.3, and 3.1.4. This special 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 described in the Annual Radiological Environmental Operating Report. The levels of naturally-occurring radionuclides which are not included in the plant's effluent releases need not be reported.
4.1 .1.2.3 If adequate samples of milk, or during the growing season, grass or leafy vegetation, can no longer be obtained from one or more of the sample locations required by Table 4-1, or if the availability is frequently or persistently wanting, efforts shall be made: to identify specific locations for obtaining suitable replacement samples; and to add any replacement locations to the REMP given in the ODCM within 30 days. The specific locations from which samples became unavailable may be deleted from the REMP. Pursuant to Technical Specification 5.5.1, documentation shall be submitted in the next Radioactive Effluent Release Report for the change(s) in the ODCM, including revised figure(s) and table(s) reflecting the changes to the location(s), with supporting information identifying the cause of the unavailability of samples and justifying the selection of any new location(s).
4.1.1.2.4 This control does not affect shutdown requirements or MODE changes.
4.1.1.3 Surveillance Requirements The REMP samples shall be collected pursuant to Table 4-1 from the locations described in Section 4.2, and shall be analyzed pursuant to the requirements of Table 4-1 and Table 4-3.
Required detection capabilities for thermoluminescent dOSimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.
Program changes may be initiated based on operational experience.
Analyses shall be performed in such a manner that the stated MINIMUM DETECTABLE CONCENTRATIONs (MDCs) will be achieved under routine conditions. Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering radionuclides, or other uncontrollable circumstances may render these MDCs unachievable. In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.
4.1.1.4 Basis The REMP required by this control provides representative measurements of radiation and of radioactive materials in those exposure pathways, and for those radionuclides, which lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation. The REMP implementsSection IV.B.2, Appendix 1,10 CFR 50, and thereby supplements the radiological effluent monitoring program by measuring concentrations of radioactive materials and levels of radiation, which may then be compared with those expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.
The detection capabilities required by Table 4-3 are within state-of-the-art for routine environmental measurements in industrial laboratories.
4-2 VER27
- -- .... ~.- ... ~.~-~
0}
Exposure Sampling and Type and c:r Number of Representative Samples and Sample (f)
Pathway Collection Frequency of
.j::>.
Locations(1 )
and/or Sample Frequency Analysis
1. DIRECT RADIATION Direct Thirty-six or more routine monitoring stations, either Quarterly. Gamma dose JJ Radiation(2) with two or more dosimeters, or with one instrument quarterly. D>
a.
for measuring and recording dose rate continuously, o*
placed as follows: o 1<0
~I An inner ring of stations, one in each meteorological m sector in the general area of the site boundary. ::l
~
o
l An outer ring of stations, one in each meteorological 3 sector at approximately 5 miles from the site. (I)
l S'
The balance of the stations to be placed in special s:
interest areas such as population centers, nearby o
l f" residences, schools, and in one or more areas to VJ o serve as control stations. ::l
<0
2. AIRBORNE Radioiodine Samples from 5 or more locations as follows: Continuous Radioiodine
'"0 IS and sampler operation Canister: 1-131 a Particulates Three or more samples from close to the three site with sample analysis weekly. 3 boundary locations, in different sectors. collection weekly, or more frequently Particulate One sample from the vicinity of a community having if required by dust Sampler: Gross the highest calculated annual average ground-level loading. beta radioactivity D/Q. analysis following filter change, and One sample from a control location, as, for example, gamma isotopic a population center 10 to 20 miles distant and in the analysis of <
m m least prevalent wind direction. composite (by G')
JJ '"0 I\) location) oo
-..J quarterly. (3)(4)
()
s:
~
0 Exposure Number of Representative Samples Sampling and Collection Type and Frequency of m Pathway and Sample Locations( 1) Frequency Analysis f" and/or Sample
3. WATERBORNE Surface(5} One sample upstream. Composite sample over Gamma isotopic analysis 1-month period.(6) monthly; composite for One sample downstream. tritium analysis Quarterly.(4) JJ Drinking Two samples at each of the one to Composite sample of river 1-131 analysis on each ~
three nearest water treatment plants water near intake at each sample when the dose o*
that could be affected by discharges water treatment plant over calculated for the g 0'
from the facility. 2-week period when 1-131 consumption of the water is ~
analysis is performed, greater than 1 mrem per m
J Two samples at a control location. monthly composite year. Composite for gross <
otherwise; and grab sample beta and gamma isotopic ~~r
J of finished water at each analyses monthly. 3 (1) water treatment plant every Composite for tritium :J
../:>. 2 weeks or monthly, as analysis quarterly.(4)(7) sr I
../:>. appropriate. (6) i s:
o Sediment from One sample from downstream area Semiannually. Gamma isotopic analysis :J Shoreline with existing or potential recreational semi-annually.(4} o value. :J Groundwater An adequate number of onsite wells Quarterly Tritium and gamma to provide detection of radioactive isotopic quarterly. Other liquid releases into the groundwater analyses based on results from tanks, underground piping, and of tritium and gamma.
other plant equipment. At least one well at a control location.
m (j)
-u
~ oo JJ I\:)
...... o s:
- - . -......... ~-.- .... ~.- ......- - . -...... Pi Exposure Sampling and Type and 0' Number of Representative Samples and Sample (j)
Pathway Collection Frequency of f' Locations( 1) and/or Sample Fre~quenc1 Analysis ~
4. INGESTION Milk Samples from milking animals in three locations within Semimonthly. Gamma 3 miles distance having the highest dose potential; if there isotopic are none, then one sample from milking animals in each of analysis semi-three areas between 3 and 5 miles distance where doses monthly. (4)(8) are calculated to be greater than 1 mrem per year.(7) ~
a.
o*
0 One sample from milking animals at a control location about a 10 miles distant or beyond, and preferably in a wind direction &i' of low prevalence. m
I Fish At least one sample of any commercially and recreation ally Semiannually. Gamma <
important species in vicinity of plant discharge area. isotopic ~r
I analyses on 3
(!)
At least one sample of any species in areas not influenced edible by plant discharge. portions.(4) S-s:
o f'
(}1 At least one sample of any anadromous species in vicinity of During spring Gamma ::::I plant discharge. spawning isotopic 5:::I.
season. analyses on ::::I edible portion.(4)
Grass or One sample from two onsite locations near the site boundary Monthly during Gamma Leafy in different sectors. growing season. isotopic. (4)(8)
Vegetation One sample from a control location about 15 miles distant. Monthly during Gamma growing season. isotopic. (4)(8)
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TABLE NOTATIONS ([)
~
I
...a.
(1) For each sample location in this table, specific parameters of distance and direction sector from a pOint midway between the center of the two reactors, and additional description where pertinent, are provided in Table 4-4, and in Figure 4-1 through Figure 4-4 of this ODCM.
n g)
(2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate Q.
continuously, ()"
0 may be used in place of or in addition to integrating dosimeters. For the purpose of this table, a o o*
thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are ~
considered as two or more dosimeters. Film badges shall not be used as dosimeters for measuring direct m
l radiation. :$.
~
m g (3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after 3
([)
sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than Br 10 times s:
o the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples. ::l S
(4) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that ::r may be attributable to the effluents from the facility.
(5) The upstream sample shall be taken at a distance beyond significant influence of the discharge. The downstream sample shall be taken in an area beyond but near the mixing zone.
(6) Composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.
< (7) The dose shall be calculated for the maximum organ and age group, using the methodology and parameters ~
m in the ODCM. "'0
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(8) If gamma isotopic analysis is not sensitive enough to meet the required MDC for 1-131, a separate analysis for 1-131 will be performed.
s:

NL-11-0811, Annual Radioactive Effluent Release Reports for 2010, Page 2-26 Through End

Text

SUMMARY

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