Text

Semiannual Radioactive Effluent Release Rept for Jul-Dec 1988
	ML20235R967
Person / Time
Site:	Fermi
Issue date:	12/31/1988
From:	Sylvia B; DETROIT EDISON CO.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	CON-NRC-89-0022, CON-NRC-89-22 NUDOCS 8903030507
	Download: ML20235R967 (198)
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{{#Wiki_filter:-._ _3-I I DETROIT EDISON COMPANY l FERMI 2 NUCLEAR POWER PLANT OPERATING LICENSE NO. NPF. - 43 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT for the period of July 1,1988 through December 31,1988 i 8903030507 881231 DR ADOCK 0500 41 g)

l-Efflu:nt R:l:aso R:part l F;bruary 1989 L TABLE OF CONTENTS PAGE 1 1. Introduction 2 2. Regulatory Limitt 3 3. Maximum Permissible Concentration 4 4. Average Energy 4 5. Measurements and Approximations of Total Activity 7 6. Abnormal Releases 8 7. Batch Relcases 9 8. Liquid Effluent Summary 11 9. Gaseous Effluent Summary 14 10. Solid Waste and irradiated Fuel Shipments '16 11. Radiological impact on Man 17 12. Radiation instrumentation 18 13. Meteorological Data Summary 18 14. Changes to the Process Control Program (PCP) 18 15. Changes to Dose Caciculation and Environmental Monitoring Locations 18 16. Changes to the Offsite Dose Calculation Manual (ODCM) Appendix A Meteorological Date Tables Appendix B Revised Offsite Dose Calculation Manual (ODCM) .______________________________.-m_________.__

Efflu nt R:l:asa R: port February 1989 Paga 1 The Detroit Edision Fermi 2 Nuclear Power Plant is designed and operated to strictly control and monitor the release of radioactive effluents to the environment in accordance with Nuclear Regulatory Commission (NRC) and Detroit Edison Company requirements. This Semlannual Radioactive Effluent Release Report is submitted in accordance with Fermi-2 Technical Specification 6.9.1.8 and NRC Regulatory Guide 1.21. This report provides the following information required by those references: 1. Summation of the quantities of radioactive material (in the form of gases and liquids) released from the plant and analysis of the radiological impact of these releases 2. Summation of quantities of radioactive material contained in solid waste packaged and shipped for off-site disposal 3. Changes to the Process Control Program (PCP) 4. Changes to the Offsite Dose Calculation Manual (ODCM) 5. Meteorological Data This report covers the period of July 1 to December 31,1988. During 1988, the total gaseous and liquid radioactive effluent releases and resulting dose to the public were maintained As Low As Reasonably Achievable (ALARA). A summary of the dose due to radioactive effluents in comparison to NRC limits is shown below: NRC DOSE UMITS FERMI-2 ESTIMATED DOSE (10CFR50 APPENDIX 1) in 1988 GASEOUS EFFLUENTS Noble Gases (Unrestricted /trea) $10 gamma mrad / year to air 1.79 E-4 mrad $20 beta mrad / year to air 2.30 E-4 mrad Dose to an individual from I-131,133, Tritium and Particulate } $15 mrem / year to any organ 1.20 E-2 mrem LIQUID EFFLUENTS $3 mrem / year to the total body 4.15 E-3 mrem $10 mrem / year to any organ 2.52 E-2 mrem Section 11 of this report presents the supporting data behind the summation. l -____________________ _ ____ w

Efflusnt Rtleass Rzport February 1989 Page 2 2. REGULATORY LIMITS The Nuclear Regulatory Commission limits on liquid and gaseous effluents are incorporated in the Fermi 2 Technical Specifications. These limits proscribe the maximum quantities and rates of release for radioactive effluents resulting from normal operation'of Fermi-2. The limits are defined in several ways to limit the overallimpact on persons living near the plant. The limits are described below: A. Gaseous Effluents 1. Dose rate due to radioactive materials releat,ed in gaseous effluents from the site to areas at and beyond the site boundary shall be limited to the following: a. Noble gases Less than or equel to 500 mrem / year to the total body Less than or equal to 3000 mrem / year to the skin b. lodine 131,133, tritium, and for all radionuclides in particulate form with - half lives greater than 8 days. Less than or equel to 1500 mrem / year to any organ. 2. Air dose due to noble gases released in gassous effluents from the reactor to areas at and beyond the site boundary shall bo limited to the following: a. Less than or equal to 5 mrads for gamma radiation Less than or equal to 10 mrads for beta radiation -During any cal 6ndar quarter b. Less than or equal to 10 mrads for gamma radiation Less than or equal to 20 mrads for beta radiation -During any calendar year 3. Dose to a member of the public from lodine-131,133, tritium, and all radionuclides in particulate form with half lives greater than 8 days in gaseous effluents released from the reactor to areas at and beyond the site boundary shall be limited to the following: a. Less than or equal to 7.5 mrems to any organ -During any calendar quarter b. Less than or equal to 15 mrems to any organ -During any calendar year 8. Liquid Effluents 1. The concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited to the concentrations specified in Title 10 of the Code of Federal Regulations Part 20 (Standards for Protection Against Radiation), Appendix B, Table ll, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases,

Efflugnt R l:ase R: port F:bruary 1989 Page 3 the concentration shall be limited to 2x10-4 (.0002) rricrocuries/ml total activity. 2. The dose or dose commitment to a member of the public from radioactive materials in liquid effluents released from the reactor to unrestricted areas shall be limited to: a. Less than or equal to 1.5 mrem to the total body Less than or equal to 5 mrem to any organ -During any calender quarter b. Less than or equal to 3 mrem to the total body Less than or equal to 10 mrem to any organ -During any calender year 3. MAXIMUM PERMISSIBLE CONCENTRATION (MPC) As required by NRC Regulatory Guide 1.21, the MPC's used to calculate permissible release rates and concentrations are described below: A. Gases The dose rate due to gaseous effluents is calculated in accordance with the Fermi 2 Offsite Dose Calculation Mancal (ODCM). The maximum permissible dose rates for gaseous releases are defined in Fermi 2 Technical Specificat!ons: Technical Specification 3.11.2.1.a (Dose rate at tho s!te boundary from gaseous effluents in the the form of noble gases): -Less than or equal to 500 mrem / year to the total body -Less than or equal to 3000 mrem / year to the skin Technical Specification 3.11.2.1.b (lodine-131,133, tritium and particulate with half-lives greater than 8 days): I -Less than or equal to 1500 c.irem/ year to any organ B. Liquids Allowable liquid releass rates are caiculated in accordance with the Fermi-2 Offsite Dose Calculation Manual ('ODCMt The maximum permissible concentration (MPC) for liquios used for these calculations are taken from 10CFR20, Appendix B, Table 11, Column 2. The most restrictive MPC is used in all cases. For dissolved and entrained gases the MPC of 2E-4 microcurios/mi is applied. This MPC is based on the Xo-135 j MPC in air (submersion dose) converted to an equiva:ent concentration in water as l discussed in the International Commission on Radiological Protection (ICRP) i Pubhcation 2. 1

Efflunnt R:l:aso RIport Fcbru ry 1989 Page 4 4. AVERAGE ENERGY The Fermi 2 Technical Specifications limit the site boundary dose rates for fission and activation gases to less than or equal to 500 mrem /yr to the total body and less than or equal to 3000 mrem /yr to the skin. Therefore, the average beta and gamma o energies (E) for gaseous effluents required to ba reported by Regulatory Guide 1.21 are not applicable to Fermi 2 and need not be reported. 5. MEASUREMENTS AND APPROXIMATIONS OF TOTAL ACTIVITY As required by NRC Regulatory Guide 1.21, this section describes the methods used to measure the total radioactivity in effluent releases and to estimate the overall errors associated with these measurements. The effluent monitoring systems are described in Chapter 11.4 of the Fermi-2 Updated Final Safety Analysis Report (UFSAR). A. Gaseous Effluents 1. Fission and Activation Gases Samples are obtained from each of the seven plant radiation monitors which continuously monitor the six ventilation exhaust points. The fission and activation gases are quantified by gamma spectroscopy analysis of periodic samples. The following are typical fission and activation gases that are quantified for dose calculations: Krypton (Kr)-87 Krypton (Kr)-88 Xenon (Xe)-133 Xenon (Xe)-133m Xenon (Xe)-135 Xenon (Xe)-138 The values reported in Section 9 are the sums of all fission and activation gases quantified at all monitored release points. Considering the inherent variability in radiation measurement and the uncertainties in sample volume, flow rate, and pressure measurements, Detroit Edison estimates that the total uncertainty of its fission and activation gas measurements is 7 percent low and 50 percent high. 2. Radiciodines Samples are obtained from each of the seven plant radiation monitors, which continuously monitor the six ventilation exhaust points. The radiciodines are entrained on charcoal and then quantified by gamma spectroscopy analysis. For each sample the duration of sampling and continuous flow rate through the charcoal are used in determining the concentration of radioiodinn. From the flow rate of the ventilation system a rate of release can be determined. The radiciodines usually quantified for dose calculations are the following: lodine (1)-131 lodine (l)-132 lodme (1)-133 lodine (1)-135 The values reported in Section 9 are the sums of all radiciodines quantified at all continuously monitored release points.

[:. } Efflurnt R ltasa R port l Fcbruary 1989 l Page 5 Considering the inherent variability in radiation measurements and the uncertainties in sample volume, flow rate, and pressure measurements, Detroit Edison estimates that the total uncertainty of these measurements is 23 percent low and 55 percent high. 3. Particulate Samples are obtained from each of the seven plant effluent radiation monitors, which continuously monitor the six ventilation exhaust points. The particulate are collected on a filter and then quantified by gamma spectroscopy analysis. For each sample the duration of sampling and continuous flow rate through the filter are used in determining the concentration of particulate. From the flow rate of the ventilation system a rate of release can be determined. The particulate usually quantified for dose calculations are the following-Manganese (Mn)-54 Iron (Fe)-59 Cobalt (Co)-58 Cobalt (Co)-60 Zinc (Zn)-65 Cerium (Ce)-144 Cesium (Cs)-134 Cesium (Cs)-137 Cerium (Ce) 141 Molybdenum (Mo)-99 (Also other quantified radionuclides with ha!? lives greater than 8 days) A composite of the filters from each ventilation release point are analyzed monthly for gross alpha radioactivity using gas proprotional counting methods. Quarterly the filters are radiochemically separated and analyzed for Strontium (Sr) 89/90 by using gas proportional counting. If found these radionuclides are reported as total particulate activity. The values reported in Section 9 are the sums of all particulate quantified at all monitored release points. Considering the inherent variability in radiation measurements and the uncertainties in sample volume, flow rate, and pressure measurements, Detroit Edison estimates that the total uncertainty of these measurements is 23 percent low and 55 percent high. 4. Tritium Samples are obtained for each of the seven plant effluent radiation monitors which continuously monitor the six ventilation exhaust points. The sample is passed through a bottle containing water and the tritium is " washed" out to the collecting water. Portions of the collecting water are analyzed for tritium esing liquid scintillation counting techniques. For each sample, the duration of sample and sample flow rate is used to determine the concentration. From the flow rate of the ventilation system a release rate can be determined. The values reported in Section 9 are the sums of all tritium quantified at all monitored release points.

c_ Efflu nt R:l:asa R: port F:bruary 1989 Paga 6 Considering the ir;terent variability in radiation measurement and the uncertainties in sample volume, flow rate, and pressure measurements, Detroit Edison estimates that the total uncertainty of these measurements is 12 percent low and 51 percent high. B. Liquid Effluents The liquid radwaste processing system and the liquid effluent monitoring system are described in the Fermi-2 UFSAR. 1. Fission and activation products Before the contents of each holding tank is discharged to the environment, a representative sample of the tanks contents is taken and retained. This sample is representative of the tanks contents. The sample allows for the determination of radioactive material concentrations and establishes the rate at which the radioactive material can be discharged to the environment. Radioactive activation and fission products that are typically found include the following: Manganese (Mn)-54 fron (Fe)-59 Cerium (Ce-144) Cobalt (Co)-58 Cobalt (Co)-60 Zinc (Zn)-65 Molybdenum (Mo)-99 At the end of the calendar quarter a composite sample is made of all discharge samples taken during the quarter. This composite sample consists of portions of each discharge sample which are proportional to the volumes discharged. The composite sample is analyzed for Iron (Fe)-55 and Strontium (Sr)-89/90. Both analytical methods involve radiochemical separation. Quantification of Fe-55 is by liquid scintillation counting and quantification of Sr-89/90 is by gas proportional counting. The values reported in Section 8 are the sums of all fission and activation products found in all batch releases. I Considering the inherent variability in radiation measurement and the uncertainties in sample flow rate and volume measurements, Detroit Edison l estimates that the total uncertainty in liquid fission and activation product l measurements is less than 14 percent. 2. Tritium l l Before the contents of each holding tank is discharged to the environment, a representative sample of the tank contents is taken and retained. At the end of the calendar month a composite sample is made of all discharge samples taken during the month. This composite sample consists of portions of each discherge sample whict, are proportional to the volumes discharged. The 4 composite sample is analyzed for tritium by liquid scintillation counting. l The values reported in Section 8 sums all tritium quantified from all batch releases. l l l _ _ __ ____ - - _ _ O

Efflu nt Riliass R: port F:bruary 1989 Page 7 Considering the inherent variability in radiation measurement and the uncertainties in flow rate and volume measurement, Detroit Edison estimates the total uncertainty in Tritium measurements is less than 14 percent. 3. Dissolved and Entrained Gases Prior to releasing liquid radioactive waste to the environment a sample is taken from the radwaste holding tank. This sample is representative of the tanks contents. The sample is examined using gamma spectroscopy to determine the dissolved and entrained noble gases. The following radiogases are typical of those which may be found: Krypton (Kr)-85m Xenon (Xe)-131 Krypton (Kr)-85 Xenon (Xe)-133 Krypton (Kr)-88 Xenon (Xe)-137 Krypton (Kr)-89 Xenon (Xe)-138 The values reported in Section 8 are the sums of all radiogases found for all batch releases. Considering the inherent variability in radiation measurement and the uncertainties in flow rate and volume measurements, Detroit Edison estimates that the total uncertainty in dissolved and entrained gases measurements is less than 15 percent. 4. Gross Alpha Before the contents of each holding tank is discharged to the environment, a representative sample of the tank contents is taken and retained. At the end of the calendar month a composite sample is made of all discharge samples taken during the month. This composite sample consists of portions of each discharge sample which are proportional to the volumes discharged. The composite sample is analyzed for gross alpha radioactivity by gas proportional counting. The values reported in Section 8 are the sums of the gross alpha radioactivity from all batch releases. Considering the inherent variability in radiation measurement and the uncertainty in flow rate and volume measurements, Detroit Edison estimates that the total uncertainty in liquid gross alpha activity. measurements is less than 23 percent. 6. ABNORMAL RELEASES For the purpose of this report, an abnormal relaase is any release of radioactive material not performed in accordance the Fermi 2 license and implementing procedures. No abnormal releases occurred during the reporting period.

l j Efflu:nt R;l: ass R: port February 1989 l Page 8 1 / 7. BATCH RELEASES As required by Regulatory Guide 1.21, a summary of data for batch releases is ) provided below. The following batch liquid releases from radwaste holding tanks to I the Circulating Water Decant Line occurred between July 1,1988 and December 31, I 1988: Number of releases: 7 Total time for all releases: 2829 minutes Maximum time for a release: 564 minutes Average time for a release: 404 minutes Minimum time for a release: 159 minutes The only batch gaseous releases from Fermi 2 are the venting or purging of the primary containment (drywell) atmosphere. Since these releases pass through the reactor building ventilation or standby gas treament system and are monitored by the final effluent monitors for these pathways, separate data on these releases are not given in this report. l

E.J.: il Efflurnt R:Irass RIport February 1989 Pagn 9 i l 8. LIQUID EFFLUENT

SUMMARY

q l. - REPORT CATEGORY

SEMIANNUAL SUMMMATION OF ALL RELEASES BY QUARTER TYPE OF ACTIVITY

ALL LIQUID EFFLUENTS REPORTING PERIOD

. QUARTER 3 AND QUARTER 4

UNIT

- QUARTER 3

QUARTER 4 TYPE OF EFFLUENT A. FISSION AND ACTIVATION PRODUCTS

1. TOTAL RELEASE (NOT INCLUDING TRITIUM, GASES, ALPHA)

CURIES 2.02E-02

1.21E-02

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD

uCi/mi 2.03E-09

1.21E-09 1

B. TRITIUM

1. TOTAL RELEASE CURIES 1.43 E-01 2.83E-01 1

2. AVERAGE DILUTED CONCENTRATION DURING PERIOD uCl/mi 1.43E-08 2.83E j C. DISSOLVED AND ENTRAINED GASES

1. TOTAL RELEASE CURIES

7.01E-05

2.60E-05 l

2. AVERAGE DILUTED CONCENTRATION DURING PEKiOD

uCl/mi 7.02E-12 2.60E-12 i

D. GROSS ALPHA RADIOACTIVITY

1. TOTAL RELEASE CURIES 2.89E-00

0.00E-01 E. WASTE VOL RELEASED (PRE-DILUTION)

UTERS 1.57 E+05

2.50E+05 F. TOTAL VOLUME DILUTION 4

DISCHARGED LITERS 9.98E+09 1.00E+ 10 l

Efflu:nt Ril2ase Rsport - February 1989 Page 10 8. LIQUID EFFLUENT

SUMMARY

(continued) 1 REPORT CATEGORY

SEMIANNUAL LIQUID BATCH RELEASES TYPE OF ACTIVITY

TOTALS FOR EACH NUCLIDE RELEASED

ALL RADIONUCLIDES REPORTING PERIOD

QUARTER 3 AND QUARTER 4 BATCH RELEASES

. UNIT - QUARTER 3 QUARTER 4 NUCLIDE ALL NUCLIDES H-3

CURIES

. 1.43E-01

2.83E-01 Na-24

CURIES 2.64E-04 2.23E-03 Cr-51

CURIES

1.01E-02

5.47E-03 Mn-54

CURIES

5.36E-04

2.90E-04 Fe-59 CURIES

2.38E-05 Co-58

. CURIES 1.19E-03 7.96E-04 Co-60

CURIES

2.84E-04

2.377.-04 Zn-65

CURIES

4.23E-04 3.07 E-04 Br-83

CURIES 1.86E-03 Mo-99

. CURIES

8.54E-04

5.89E-04 Ru-103

CURIES 2.80E-04 Tc-99M CURIES 9.27E-04

, 9.96E-04 1-133

CURIES

2.69E-05 4.15E-05 Ce-144 CURIES

4.63E-04 Xe-133

CURIES 1.64E-05 X6-135

CURIES

7.01E-05 9.59E-06 Ag-110M

CURIES

1.18E-05 Sr-89

CURIES

2.36E-05

5.75E-05 Sr-90

CURIES 7.54E-07 Ba-131

. CURIES . 5.63E-05

2.02E-04 Ea-135M CURIES

9.23E-05 Re-188 CURIES 6.54E-05 As-76

CURIES

7.44E-04 Cs-134

CURIES
< 0.1 E-07
< 0.1 E-07 Cs-137

CURIES

< 0.2E-07

< 0.2 E-07 Ce-141 CURIES

< 0.2 E-07

< 0.2 E-07 Other

. CURIES

2.95E-03 (except gross alpha)

Total for Period CURIES 1.63E-01 2.95E-01 Less than Lower Limit of detection (LLD), i.e., the maximum sensitivity of measurement, in units of microcuries per milliliters (uCi/ml). ._-_____--_-__________a

Efflugnt R11: ass R: port. Fabruary 1989 Page 11 9. GASEOUS EFFLUENT

SUMMARY

REPORT CATEGORY SEMIANNUAL SUMMMATION OF ALL RELEASES BY OUARTER l TYPE OF ACTIVITY ALL AIRBORNE EFFLUENTS REPORTING PERIOD - QUARTER 3 AND QUARTER 4

UNIT

QUARTER 3

QUARTER 4 TYPE OF EFFLUENT A. FISSION AND ACTIVATION GASES 1.' TOTAL RELEASE CURIES

6.84E-01

0.00E-01

2. AVERAGE RELEASE RATE FOR PERIOD uCl/sec 8.61 E-02 0.00E B. RADIOIODINES

1. TOTAL IODINE - 131 CURIES 6.97E-05

3.34E-04

2. AVERAGE R'ELEASE RATE FOR PERIOD uCl/sec

8.77E-06

4.20E-05 C. PARTICULATE

1. PARTICULATE (HALF-UVES>8 DAYS)

CURIES

9.64E-05 1.18E-03

2. AVERAGE RELEASE RATE FOR PERIOD uCl/sec

. 1.21 E-05 1.48E-04

3. GROSS ALPHA RADIOACTIVITY CURIES 2.99E-07 0.00E-01 D. TRITIUM

1. TOTAL RELEASE CURIES 0.00E-01 0.00E-01

2. AVERAGE RELEASE RATE FOR PERIOD uCl/sec 0.00E-01

0.00E-01

Efflutnt RIl ass R: port Fsbruary 1989 P:ga 12 9. GASEOUS EFFLUENT

SUMMARY

(continued) l REPORT CATEGORY SEMIANNUAL AIRBORNE CONTINUOUS RELEASES TYPE OF ACTIVITY

FISSION GASES, IODINES, AND PARTICULATE REPORTING PERIOD QUARTER 3 AND QUARTER 4 GROUND RELEASES

UNIT

. QUARTER 3 . QUARTER 4 NUCLIDE PARTICULATE Cr-51

CURIES

. 8.57E-06 . 6.47E-04 Mn-54

CURIES

2.07E-05

2.33E-05 Co-58

CURIES

2.54E-05 6.84E-05 Mo-99

CURIES

0.00E+00

1.53E-04 Ba-140

. CURIES

0.00E+00

. 9.54E-05 La-140

CURIES 0.00E+00

2.69E-04 Ce-144

. CURIES

0.00E+00 6.74E-06 Co-57 CURIES

0.00E+00

2.32E-06 Tc-99m

. CURIES

9.54E-05

5.76E-04 Ba-139

CURIES

0.00E+00 1.51 E-03 Y-91m4

CURIES

0.00E+00

1.62E-04 Sr-91 CURIES

0.00E+00

2.02E-04 Zn-65 CURIES

. 0.00E+00

4.76E-05 Co-60

CURIES

0.00E+00

1.61E-05 Na-24

CURIES

. 0.00E+00

2.71E-02 Ru-103

CURIES

1.51E-05

. 1.31 E-05 Ba-131 CURIES .0.00E+00

1.64E-04 Ba-135m

. CURIES

0.00E+00 8.60E-04 As-76 CURIES 0.00E+00

5.07E-04 Rb-89

. CURIES

0.00E+00

3.54E-04 Cs-138 CURIES 0.00E+00

1.78E-04 Sr-89

. CURIES 1.64E-05 8.00E-05 Sr-90 CURIES

1.02E-05

. 1.81 E-05 Fe-59 . CURIES

< 0.1 E-11
< 0.1 E-11 Cs-134 CURIES
< 0.1 E-11
< 0.1 E-11 Cs-137 CURIES
< 0.1 E-11

< 0.1 E-11 Ce-141

CURIES

< 0.1 E-11

< 0.1 E-1 1 Ce-144 CURIES

"" < 0.1 E-1 1

< 0.1 E-11 Total for Period

CURIES 1.92E-04

. 3.31E-02 Less than Lower Umit of Detection (LLD), i.e., the maximum sensitivity of measurement in units of microcuries per milliliter (uCl/ml), i i ___j

Efflusnt R:lrase R7 port j Fcbruary 1989 - Page 13 '9. GASEOUS EFFLUENT

SUMMARY

(continued) REPORT CATEGORY.

SEMIANNUAL AIRBORNE CONTINUOUS RELEASES TYPE OF ACTIVITY -

FISSION GASES, IODINES, AND PARTICULATE REPORTING PERIOD

QUARTER 3 AND QUARTER 4 GROUND RELEASES

.: UNIT

QUARTER 3

. QUARTER 4 NUCLIDE FISSION GASES Xe-135 CURIES

6.84E-01

0.00E-01 Kr-87 CURIES '

< 0.1 E-04

< 0.1 E-04 Kr-88

CURIES

< 0.1 E-04
< 0.1 E-04 Xe-133 CURIES
< 0.1 E-04
< 0.1 E-04 Xe-133m

CURIES

< 0.1 E-04
< 0.1 E-04.

Xe-138

CURIES

< 0.1 E-04

< 0.1 E-04 TOTAL FOR PERIOD CURIES 6.84E 0.00E-01 IODINES l-131 CURIES

6.97E-05

3.34E-04 l-132

CURIES

. 0.00E+00 1.01E-05 ' l-133

CURIES

. 3.73E-04

6.02E-03 1-135 CURIES

. 0.00E+00 8.44E-05 Total for Period CURIES

4.43E-04

6.45E-03 l

J

Efflu:nt Rslaass R: port Fsbruary 1989 ^ Page 14 10. SOLID WASTE AND 1RRADIATED FUEL SHIPMENTS A. Solid Waste Shipped Offsite for burial or disposal (not irradiated fuel) Container 6 month Est. Total 1 1. Type of Waste Volume Unit period Error % 3 a. Spent resins, filter sludges, m 8.50E+01 +25 evaporator bottoms, etc. Curies 4.29E+00 225 3 b. Dry compressible waste, m 4.27E+01 +25 contaminated equipment, etc. Curies 1.19E+00 25 c. Irradlated components, control rods, etc. O d. Other 0 2. Estimate of major nuclide composition (by type of waste) a. Spent resins, filter sludges, evaporator bottoms, etc. Percent of Nuclide Total Activity Curles Cr-51 40.8 1.75E+02 Mn-54 7.6 3.27E+01 Fe-55 20.8 8.91 E+01 Co-58 8.2 3.53E+01 Co-60 5.4 2.31 E+01 Fe-59 3.0 1.28E+01 Zn-65 6.2 2.64E+01 H-3 < 0.1 2.98E-02 C-14 0.5 2.16E+00 Zr-95 <0.1 1.38E-02 Ba-131 4.9 2.10E+01 Ce-144 2.5 1.08E+01 Sr-90 <0.1 6.79E-04 Ni--83 0.1 4.20E-01 Cs-137 <0.1 6.26E-02

Efflu:nt Ralzase R! port Fzbruary 1989 Page 15 b. Dry compressible waste, contaminated equipment, etc. Cr-51 12.3 1.45E-01 Mn-54 15.8 1.88E-01 Fe-55 25.6 3.05E-01 Co-58 18.0 2.14E-01 Co-60 15.8 1.88E Zn-65 6.4 7.65E-02 Ni-63 1.2 1.48E-02 C-14 4.9 5.80E-02 Note: Activities of all principal radionuclides were determined hy measurement. 3.- Solid Waste Disposition (All Class A waste shipped in LSA containers) Type of shipment /. Number of Mode of solidification process shipments Transport. Destination Dewatered resin ' 9 truck Barnwell, SC Cement solidified powder resin / filter sludge 2 truck Barnwell; SC Dry active waste 2 truck Channahon, IL Contaminated equip. 1 truck Wampum, PA l 4. Irradiated Fuel Shipments: None i

Efflu nt R lease R port F:bruIry 1989 Page 16 11. RADIOLOGICAL IMPACT ON MAN A. Dose Due to Liquid Effluents As discussed in Section 2.5.1, the Fermi 2 Offsite Dose Calculation Manual (see Section 1b of this report) the maximum potential dose to an individual due to liquid effluents is based on the combined pathways of fish consumption and water consumption. The following are the maximum individual organ doses for all of 1988 calculated according to Section 2.5.1 of the ODCM: Organ 1988 Liquid Effluent Dose Bone 3.21E-3 mrem Liver 9.18E-3 mrem Total Body 4.15E-3 mrem Thyroid 1.77E-4 mrem Kidney 5.69E-3 mrem Lung 1.61E-4 mrem Gl/LLI 2.52E-2 mrem B. Dose Due to Gaseous Effluents As discussed in Section 3.8.1 of the Fermi 2 Offsite Dose Calculation Manual (see Section 16 of this report), the maximum potential dose to an individual due to gaseous effluents is based on the identification of the individual who will be maximally exposed by the inhalation, ingestion, and ground plane pathways. The following are the maximum individual organ doses for all of 1988 calculated according to Section 3.8.1 of the ODCM. Organ 1988 Gaseous Effluent Dose Bone 3.93E-4 mrem Liver 2.63E-4 mrem Thyroid 1.20E-2 mrem Kidney 2.06E-4 mrem Lung 1.12E-4 mrem GI/LLI 2.11E-4 mrem Total body 2.15E-4 mrem C. Dose Due to Direct Radiation and Compliance with 40CFR190 Title 40, Part 190 of the Code of Federal Regulations requires that dose to an individual from the uranium fuel cycle be limited to 25 mrem /yr to the total body and 75 mrem /yr to the thyroid. The sources of fuel cycle dose not j analyzed in Sections A and B above are due to other fuel cycle tacilities rend dose due to direct radiation. As discussed in Section 4.2 of the Fermi 2 Offsite Do,e Calculation Manual (see Section 16 of this report), no othei fuel cycle facilities contribute significantly to dose in the vicinity of Fermi 2. With respect to direct radiation, none of the offsite TLD locattons listed in Table 6.0-1 of the ODCM showed 1983 TLD readings which were significantly greater than the TLD readings at the control locations. Since other facilities and direct radiation did

1 Efflu:nt R;liaso R: port February 1989 Page 17 I not contribute significantly to offsite dose, and since Sections A and B above show compliance with the more restrictive requirements of 10CFR50 Appendix 1, Fermi 2 was in compliance with 40CFR190 in 1988. D. Dose to Visitors on Site As discussed in Section 4.0 of the Fermi 2 Offsite Dose Calculation Manual, " visitors" to the Fermi 2 site may receive dose due to their activities within the site boundary. For purposes of this analysis, visitors are members of the public who spend time within the site boundary, and whose work is not associated with the operation of Fermi 2. The ODCM considers two categories of visitors: Persons ice-fishing on Lake Erie and persons spending time in the Fermi 2 Visitors' Center. The ODCM lists the maximum amount of time an individual is likely to spend in these activities, and the dispersion factors and exposure pathways which apply: Exposure by direct radiation from noble gases and by inhalation of radioactive particulate, lodines, and tritium are considered. (These pathways are in addition to those already considered, such as fish consumption in the case of ice fishermen.) Based on the above assumptions, the maximum dose in 1988 to an individual as a result of ice fishing within the site boundary is 5.97E-6 mrem to the total body and 6.61E-4 mrem to the maximally exposed organ (thyroid). This is in addition to previously analyzed doses due to fish consumption. The maximum dose in 1988 to a visitor at the Visitors' Center is 1.73E-7 mrem to the total body and 1.88E-6 to the maximally exposed organ (thyroid). 12. RADIATION INSTRUMENTATION Fermi 2 Technical Specifications 3.3.7.11, Radioactive Liquid Effluent Monitoring Instrumentation, and 3.3.7.12, Radioactive Gaseous Effluent Monitoring instrumentation, require that those monitors which exceed the time specified for out of service be reported in the next Semlannual Effluent Release Report. During this reporting period, July through December of 1988, the time specified in the action statements for these monitors was not exceeded. I

Efflu:nt P.algass Report February 1989 Page 18 13. METEOROLOGICAL DATA

SUMMARY

The meteorological monitoring system is described in the Fermi 2 UFSAR. In accordance with Regulatory Guide 1.21, data recorded by that system is provided here to permit the Nuclear Regulatory Commission to assess the radiological impact of Fermi 2 releases independently. The format for the data required by Regulatory Guide 1.21 is used. Appendix A contains the meteorological data tables. 14. CHANGES TO THE PROCESS CONTROL PROGRAM (PCP) As required by the Fermi 2 license the operator (Detroit Edison) is required to establish a program that will reasonably assure the complete processing of radioactive wastes. This program assures processed wastes are completely solidified and are free of standing water. Changes to the PCP Manual are provided to document changes to pre-established conditions and to ensure that controls are in place to assure that the radioactive waste is solidified. During this reporting period, July through December of 1988, there were no changes to the PCP. 15. CHANGES TO DOSE CALCULATION AND ENVIRONMENTAL MONITORING LOCATIONS As a result of the 1988 Land Use Census, the following changes were made: A. A new critical receptor location (for calculation of maximum individual doses due to gaseous effluents) was identified: 4262 Post Road,3379 meters (2.1 miles) WNW of the plant. B. Two Food Product locations, formerly called FP-2 and FP-4, were dropped due to lack of collectible broadleaf vegetation. C. Two new Food Product locations were added: FP-4 at 4262 Post Road and FP-6 at 8200 Geirman Road as the new control location. D. One Milk location (M-4) was dropped due to the death of the milk animal. ] E. One Milk location (M-6) was added, but dropped after only 4 samples were collected because of insufficient samples. 16. CHANGES TO THE OFFSITE DOSE C'ALCULATION MANUAL (ODCM) On December 15,1988, the Fermi 2 Onsite Review Organization (OSRO) approved a completely revised ODCM. This new ODCM became effective Jcnuary 1,1989, and was used in the calculation of doses for this report. The new ODCM incorporates changes from the 1988 Land Usa Census. i I

- Efflutnt R:Irass R2 port ~ Febru:ry 1989 Page 19 The new ODCM is enclosed as Appendix B, with a copy of the Licensing Change Request presented to OSRO containing a summary of the revisions. O d i ( l;

EffluInt R;l ase Rep:rt Fcbruary 1989 APPENDIX A: METEOROLOGICAL DATA TABLES

1 I s. 'i JOINT FREQUENCY DISTRIBUTION (JFD) AT THE 10-METER LEVEL FIRST QUARTER 1988 't I . 4 li """"a*""o" o - -_ _ _

i .1 1 8 i 5 B. R !g 3 86 2 22 3 4 l-a

s 4

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=

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~~== M E E M M N 4 4 ( .J W W O=CN*COOOOO T .J 4 W 00**e000000 O U g N M M W O M M M W~~

~~== W C0099000000 O~~

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=
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Efflu;nt R;l:ase RIpert Fcbruary 1989 i f-l l APPENDIX B: REVISED OFFSITE DOSE CALCULATION MANUAL i
66 111 - 1013 l%l - lololol LCR j v Revision Page 1 of I p,................ * *
PA R T 1 : UFSAR, PLAN, OR PRDGRAM REVISIDN [ ]NA *"***""****"**""**l A) Document opF SITE Dos.c c 4 L.c u t. A m o d meal B) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) l FA n 2.E
'2.Gk:nss o4 ro mE mm M 1 C) Reason for Change h"cir e w b ric a B-socc i b c snena cF 4k e cR snt 0;ts-co \\chus re w ra Rcat ac ro L e-wt e eAen k Thw en>G Du t' 'kDE 5 hv *Ro ch ou.Not L icui1D o nc0 Sct W a r 2 M Ghe5 L w i44 T2chertuce ro % t' n G9rs p4 ;aTt. - F ti m 'l Z- "tt9 W M N t t.m Tie 4.' D) Reference end Source Documents (Identity) EDP Test PDC Tech Spec ABN Procedure DER SE (Anached) Effectiveness Review (Attached) Other Drawings. Design Calculations, Correspondence, etc, peu....u...........* * * * *
~~PART 2 : OPERATING LICENSE CHANGES (KINA ***""""*""""*"l A) Document~~
! ) Operating License [ ] Tech Specs [ ] Environmental Protection Plan I ] Tech Spec Clarification _.B) Section(s), Table (s), Figure (s), etc. Affected (Attach marked-up pages) C) Reference and Source Documents Attached [ ] Significant Hazards Consideration [ ] Environmental impact - Categorical Exclusion [ ] Environmental Evaluation [ l ] Other i D) is UFSAR change required? [ ]Yes I } No LCR No k) Priority NRC approval required by (date): An [ ) Emergency [ ) Exigent condition will occur if not approved by: (State dete): Explanation F) implementation DER No. r............................, A R T 3. A,, R D v A,.S..................................................... i 88!ik [68 [ Date A) Originator B) Technical Expert A .p,. / Date 12- [ 6 6 Date /A M C) Nuclear Organlaation Unit Head e // D) Director, Nuclear Engineering l/NA Date E) Plant Manager /) Date / 2 'f 7 / l F) Director. Nuclear Licensing /!(.h///o ,$ //-/,A e u Date /2// /ff t / s a N DSRD Approval [ ] S Date /.2r'FM l lH) NSRG Approval [dNA Date l l Form FIP-RA2-01 Att 1 P1/1092186 DTC: File:
~ EFFECTIVENESS REVIEW 16 81 - lo l"6l>l - 10l0lM l 1 Reference LCR Revision Page 1 of I p,.u......u...............**"*******" PART 1: UFSAR [ y.J N A " * " "' " " """ """ " " " " " l j A) Cuality Assurance Program ] [ ]Yes [ ] No Does the change (s) cease to satisfy the criteria of 10CFR50, Appendix B and the UFSAR program commitments previously accepted by the NRC? Provide the basis for each change on Attachment 2, Page 2.
~~8) Fire Protection Program~~
[ ]Yes [ ] No Does the change (s) significantly decrease the level of fire protection in the plant? [ ]Yes [ ] No Does the change (s) result in failure to complete Fire Protection Program approved by the NRC prior to license issue? Provide the basis for each change on Attachment 2. Page 2. P"
PART 2: RADIOLOGICAL EMERGENCY RESPONSE PREPAREDNESS PLAN [ y3N A **""""*"" *

j A) [ ] Yes [ ] No Does the change (s) decrease the effectiveness of the RERP Plan?
[ ] Yes [ ] No Does the RERP Plan, as changed, cease to meet the standards of 10CFR50.47(b) and 10CFR50 Appendix E? Provide the basis for each change on Attachment 2 Page 2. p w.. n.........~......... PART 3 : SECURITY PLANS ['X]N A """""""""""""""""""l A) Document A) [ ] Yes I ] No Does the change (s) decrease the effectiveness of the Physical Security Plan or Security Personnel Training ano Qualification Plan prepared pursuant to 10CFR50.34(c) or 10CFR73? ! ] Yes [ ] No Does the change (s) decrease the effectiveness of the first four categories of informational Background, Generic Planning Base Licensee Planning Base, and/or responsibility matrix of the Safeguards Contingency Plan prepared pursuant to 10CFR50.34(d) or 10CFR73? Provide the basis for each change on Attachment 2, Page 2. puu................... "" P ART 4 : PROCESS CONTROL PROGRAM bc]N A"""" * """""""" l A) { ]Yes [ ] No Does the change (s) reduce the overall conformance of the solidified waste product to existing criteria for solid wastes in accordance with Technical Specification 6.13? Provide the basis for each change on Attachment 2. Pa ge 2. p,.. u.... u........... * * * " * * "" a * * *
P ART 5 : ODCM I; NA"*""""*""**"""""""*""""l A) ( ) Yes [K) No Does the change (s) reduce the accuracy or reliability of the dose calculations or setpoint determinations in in accordance with Technical Specification 6.14?
~~Provide the basis for each change on Attachment 2 Page 2. ,_.........................., ART A,, ROYALS....................................................., Date i 66 A) Originator N A w Date 82, 8f~~
~~8) Technical Exp rt cf~~
.A C) Ouality Assurance For Security Plans only) e4 ' Date lD) CSRO (Not required for UFSAR-QA Changes) tJ f A Date l ( Form FIP-RA2-01 Att 2 P1/2 092188 DTC File: L_-_______._____.__
EFFECTIVENESS REVIEW DOCUMENTATION Reference LCR 1816l - l o l1 l 2.1 lo lb (*4 l Revision Page 1 of 6 D:cument BEF 5iTE Dass cut exluru,o run te a4 ( o otm ) Usted below is each change. by section and page. the reason for the change and the basis for c ncluding that the revised plan or program continues to satisfy the criteria for that plan or program. S:ction/Page Change easis A s.4 e.e A mis REvivou oF me etx.m Dh.~.nes we ewsive acu N tulccm.,4 r.- M urJ Cooc 4.,e (, ai.tvrme,- seemoa. vue-U-) C*r4 u d GN iets se eM.us ace As 4 tt~_.M. 0)mvo omm.4 4 Nghtes (go cre.19 ENC) LM C mmus E.Nt u ins (4-) McIcd Oost s i On*. M RadiotocicuS e.orosarmentg ! Assc5 m ev4T OP wm utE Ce'"" r iI mo,* im a hs rw la
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l _ _____- -Off:lts Dese Calculation Manual wu Page 0.0-1 1 i DETROIT EDISON - FERMI 2 OFFSITE DOSE CALCULATION MANUAL 4 ARuS - surORMATION SERMCES 4M Date approved: /A -/[-// Release authortred by: Change numbers incorporated: LCR 88-032-00M Rev 2 Date DSN File 1715.02 Recipient DTC TMPLAN
m I Revisisn 2 Page 0.0-2 TABLE OF CONTENTS i Page Section 1.0-1
~~1.0 INTRODUCTION~~
2.0-1 2.0 UQUID EFFLUENTS. 2.0-1 2.1 Radiation Monitoring instrumentation and Controls 2.0-1 2.1.1 Technical Specification 33.7.11 Requirement 2.0-2 2.1.2 Nor Technical Specification Monitor 2.0-2 2.2 Sampling en) Analysis of Uquid Efnuents 2.0-3 1.2.1 SATCH Releases 2.0-3 2.2.2 CONTINUOUS Releases 2.0-3 2.3 ugued Effluent Monkor Setpoints 2.0-4 2.3.1 Uquid Redweste Effluent une Monitor (D11-N007) 2.0-7 23.2 Circulating Water Reservoir Decant Line Radiation Monitor (D11-N402) 2.0-7 2.33 Generic, Conservative Alarm Setpoint for D11-N402 2.0-8 23.4 Alarm Setpoint for GSW and RHR System Radiation Monitors 2.0-8 23.5 Alarm Response - Evolusting Actus'I Release Conditions 2.0-9 23.6 Uquid Radweste Setpoint Determination With Contaminated Circulating Water Pond 2.0-9 2.4 Contaminated OSW or RHR System - Quantifying and Controlling Releases 2.0-10 2.5 Uguld Effluent Dose Calculation - 10 CFR 50 2.0-10 '2.5.1 MEMBER OF THE PUBLIC Dose - Uguld Effluents 2.0-12 2.5.2 ' Simplified Liquid Effluent Dose Calculation 2.0-13 2.8i.3 Contaminated GSW System - Dose Calculation 2.0-14 2.6 Uquid Eftluent Dose Projections 3.0-1 3.0 GASEOUS EFFLUENTS 3.0-1 3.1 Radiation Monitoring instrumentation and Controls 3.0-1 3.1.1 Effluent Monitoring - Ventilation System Releases 3.0-1 3.1.2 . Main Condenser Offgas Monitoring 3.0-2 3.13 Reactor Building Ventilation Monitors (Gulf Atomic) 3.0-2 3.2 ' Sampling and Analysis of Gaseous Effluents 3.0-2 3.2.1 Containment PURGE 3.0-2 3.2.2 Ventilation System Releases 3.0-3 33 Gaseous Effluent Monitor Setpoint Determination 3.0-3 33.1 Ventilation System Monitors 3.0-5 3.3.2 Conservative, Generic Alarm Setpoints 3.0-5 333 Gaseous Effluent Alarm Response - Evaluating Actual Release Conditions 3.0-6 3.4 Containment Drywell VENTING and PURGING 3.0-6 3.4.1 Release Rate Evaluation 3.0-7 3.4.2 Alarm Setpoint Evaluation 3.0-7 3.5 Quantifying Releases - Noble Geses, 3.0-7 3.5.1 Quantifying Releases Using SPING Noble Gas Monitor 3.0-9 3.5.2 Quantifying Release Rate and Total Releases with Monitor inoperable
E R: vision 2 Page 0.0-3 TABLE OF CONTENTS (cantinu:d) Pege Section 3.0-10 3.6 Site Boundary Dose Rate - Radiolodine end Particulate 3.0-10 , 3.6.1 Simpitfied, Dose Rate Evalust6on for Radioiodines and Particulate 3.0-11 3.7 Noble Gas Effluent Dose Calculations - 10 CFR 50 3.0-11 -3.7.1 UNRESTRICTED AREA Dose - Noble Gases 3.0-11 3.7.2 Simplified Dose Calculation for Noble Gases 3.0-12 3.8 Radiolodine and Particulate Dose Calculations - 10 CFR 50 3.0-12 3.8.1 UNRESTRICTED AREA Dose - Radiolodine and Particulate 3.0-13 3.8.2 Simplified Dose Calculation for Radiolodines and l Particulate 3.0-14 3.9 Gaseous Effluent Dose Projection 4.0-1 4.0 SPECIAL DOSE ANALYSES 4.0-1 4.1 Doses Due to. Activities inside the SITE BOUNDARY 4.0- 1 4.2 Doses to MEMBERS OF THE PUBLIC - 40 CFR 190 4.0-2 4.2.1 Effluent Dose Calculations 4.0-3 4.2.2 Direct Exposure Dose Determination 4.0-3 4.2.3 Dose Assessment Based on R' radiological Environmental Monitoring Data 5.0-1
~~5.0 ASSESSMENT~~
OF LAND USE CENSUS DATA 5.0-1 5.1 Land Use Census as Required by TS 3.12.2 5.0-3 5.2 Land Use Census to Support Realistic Dose Assessment 6.0-1 6.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM 60-1 6.1 Sampling Locations 6.0- 1 6.2 Reporting Levels 6.0-2 6.3 Interlaboratory Comparison Program APPENDICES A-1 A Evaluation of Generic Concentration Limit for Liquid Effluents 8-1 B Technical Basis for Effective Dose Facto s Uquid Effluent Releases C-1 C Technical Basis for Effective Dose Factom Gaseous Radweste Effluents TABLES 2.0-15 2.0-1 Fermi 2 Site Specific Liquid ingestion Dose Commitment Factors, A o i 2.0-17 2.0-2 Bioaccumulation Factors (BFi) 3.0-15 3.0-1 Default Noble Gas Radionuclides Distribution of Geseous Effluents 3.0-16 30-2 Generic Values for Evaluating Gaseous Release Rates and Alarm Sotpoints 3.0-17 3.0-3 Dose Factors for Noble Gases 3.0-18 3.0-4 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations 3.0-19 3.0-5 Gaseous Effluent Pathway Dose Commitment Factors
Revislan 2 Page 0.0-4 TABLE OF CONTENTS (continu:d) Section Page TABLES 4.0-7 4.0-1 Assumptions for Assessing Doses Due to Activities inside SITE BOUNDARY 4.0-8 4.0-2 Recommended Exposure Rates in Ueu of Site Specific Data 6.0-3 6.0- 1 Radiological Environmental Monitoring Program Fermi 2 Sample Locations and Associated Media 6.0-13 6.0-2 Radiological Environmental Monitoring Program, Fermi 1 Sample Locations and Associated Media A-2 A-1 Concentration Umit for Uquid Effluents from Fermi 2 B-4 B-1 Relative Dose Significance of Radionuclides in Uquid Effluents C-4 C-1 Effective Dose Factors - Noble Oss Effluents FIGURES 2.0-18 2.0-1 Liquid Radioactive Effluent Monitoring and Processing Diagram 3.0-34 3.0-1 Gaseous Radioactive Effluent Monitoring and Ventilation Systems Diagram 6.0-15 6.0-1 Radiological Environmental Monitoring Program Sampling Locations - Site Area 6.0-16 6.0-2 Radiological Environmental Monitoring Program Sampling Locations - Greater than 5 Miles 6.0-17 6.0-3 Radiological Environmental Monitoring Program Sampling Locations - within 10 Miles 6.0-18 6.0-4 Radiological Environmental Monitoring Program Sampling Locations - Site Area (Lake Erie side) 6.0-19 6.0-5 Fermi 1 Sampling Locations END OF SECTION 0.0 l
C'ucI:ar Production - Fermi 2 ODCM-1.0 Offsita Dose Calculation Manual Revi:lon 2 Page 1.0-9 MTRODUCTION
~~1.0 INTRODUCTION~~
The Fermi 2 Offsite Dose Calculation Manual (ODCM) describes the methodology end parameters used in: 1.1 Determining radioactive material release rates and cumulative releases 1.2 Calculating radioactive liquid and gaseous effluent monitoring instrumentation alarm / trip set points 1.3 Calculating the corresponding dose rates and cumulative quarterly and yearly doses. The methodology provided in this manual is acceptable for use !n demonstrating compliance with concentration limits of 10 CFR 20.106 and the cumulative dose criteria of 10 CFR 50, Appendix i and 40 CFR Ig0, and the Fermi 2 (Radiological Effluent) Technical Specifications. More conservative calculational methods and/or conditions (e.g., location and/or exposure pathways) sspected to yield higher computed doses than appropriate for the maximally exposed person may be assumed in the dose evaluations for controlling the release of radioactive material feom Formi 2. The ODCM will be maintained at Fermi 2 for use as a reference guide and training document of accepted methodologies and calculations. Changes to the ODCM calculational methodologies and parameters will be made as necessary to ensure reasonable f conservatism in keeping with the principles of 10 CFR 50.36a and Appendix I for demonstrating radioactive effluents are "As Low As Reasonably Achievable." NOTE: Throughout this document words appearing all capitalized denote either definitions specified in the formi 2 Technical Specifications or common acronyms. I Section 2.0 of the ODCM describes equipment for monitoring and controlling liquid effluents, sampling requirements, and dose evaluation methods. Section 3.0 provides similar information on gaseous effluent controls, samplin2, and dose evaluation. Section 4.0 describes special dose analyses required for compliance with Fermi 2 Technical Specifications and 40 CFR Ig0. Section 5.0 describes the role of the annual land use census in identifying the controlling pathways and locations of exposure for assessing l potential off-site doses. Section 6.0 describes the Radiological Environmental Monitoring l Program. END OF SECTION 1.0 f ARMS - INFORMATION SERVICES Date approved: Release authortred by: Change numbers incorporated: LCR 88-032-ODM DSN Rev 2 Date DTC TMPl.AN File 1715.02 Recipient L n
C:uelear Production - Formi 2 ODCM-2.0 Offsite Dre Calculation Manual Revisisn 2 Page 2.0-1 LIQUID EFFLUENTS 2.0 UQUID EFFLUENTS This section summarizes information on the Siquid effluent radiation monitoring instrumentation and controls. More detailed information is provided in the Fermi 2 UFSAR and Fermi 2 design drawings from which this summary was derived. This section also describes the sampling and analysis required by Technical Specifications. Methods for calculating alarm setpoints for the liquid effluent monitors are presented. Also, methods for ovaluating doses from liquid effluents are provided. 2.1 Radiation Monitoring instrumentation and Controls This section summarizes the instrumentation and controls monitoring liquid effluents. This discussion focuses on the role of this equipment in assuring compliance with the Fermi 2 Technical Specifications and ODCM. 2.1.1 Technical Specification (TS) 3.3.7.11 Requirement Fermi 2 TS 3.3.7.11 prescribes the monitoring required during liquid releases and the backup sampling required when monitors are inoperable. The liquid effluent monitoring instrumentation for controlling and monitoring radioactive effluents in accordance with the Fermi 2 TS 3.3.7.11 is summarized below: 1. Radiation Alarm - Automatic Release Termination j a. Liquid Radweste Effluent Line - The D11-N007 Radiation Monitor on the liquid radweste effluent line provides the alarm and automatic termination of liquid radioactive material releases prior to exceeding 1 Maximum Permissible Concentration (MPC) (10 CFR 20, Appendix B. Table 11, Column 2) required by TS 3.3.7.11. The monitor is located upstream of the isolation Valve (011-F733) on liquid radwaste discharge line and monitors the concentration of liquid effluent before dilution by the circulating water reservoir (CWR) decent flow. 2. Radiation Alarm (only) a. Circulating Water Reservoir (CWR) Decent Une - The CWR Decent Line Radiation Monitor (D11-N402) provides Indication of the concentration of radioactive materialin the diluted radioactive liquid releases just before discharge to Lake Erie. As required by TS 3.3.7.11, the alarm setpoint is established to alarm (only) prior to exceeding MPC. ARMS - INFORMATION SERVICES Date approved: Release authorized by: Change numbers incorporatodi LCR 88-032-ODM DSN Rev 2 Date DTC TMPLAN File 1715.02 Recipient
Page 2.0-2 3. Flow Rate M::asuring Devicss a. Liquid Radweste Effluent Line - In accordance with TS 3.3.7.11, the '{ release rate of liquid redweste discharges is monitored by ') 011-R703. This flow rate instrumentation is located on the radweste discharge line prior to the junction with the CWR decent line. b. Circulating Water Reservoir Decent Line - in accordance w;th TS 3.3.7.11, the flow rate of the CWR decent line is monitored by = N71-R802. The flow rate instrumentation is located on the decant: line downstream of the junction with the liquid redweste effluent line. This instrumentation measures the total discharge flow rate from Fermi 2 to Lake Erie. 2.1.2 Non Technical Specification Monitor An additional monitor not required by Fermi 2 TS is provided by Detroit Edison to reduce the likelihood of an unmonitored release of radioactive liquids. 1. General Service Water - The General Service Water (OSW) Radiation i Monitor (011-N008) provides additional control of potential radioactive effluents. D11-N008 monitors the OSW System prior to discharge into the Main Condenser circulating water discharge line to the Circulating-Water Reservoir. Although not a TS required monitor, D11-N008 monitors a primary liquid stream in the plant that also discharges to the environment (Lake Erie via the Circulating Water Reservoir). Indication of radioactive material contamination in the OSW System I would also indicate potential CWR contamination and the need to control all discharges from the CWR as radioactive effluents. 2.2 Sampling and Analysis of Liquid Effluents The program for sampling and analysis of liquid waste is prescribed in the Fermi 2 Technical Specifications, Table 4.11.1.1.1-1. This table distinguishes two types of liquid releases: 2.2.1 BATCH releases, defined as discrete volumes, normally processed through the radweste system to the waste sample tanks 2.2.2 CONTINUOUS releases, from the Circulating Water Reservoir (CRW) System, if it becomes contaminated Continuous releases from the CWR System are via the CWR decent line to Lake Erie. The CWR System is not expected to become contaminated. Therefore. continuous l radioactive material releases are not expected. However, the General Service Water l (GSW) and the CWR systems interface with radioactive systems in the plant. Also, the OSW intake is within a few hundred feet of the CWR decent line discharge to l Lake Erie. For these reasons, it is prudent to consider the GSW and the CWR a potential source of radioactive effluents and to sample them regularly. i
ODCM-2.0 Revisien 2 Page 2.0-3 2.2.1 BATCH Releases Fermi 2 TS Table 4.11.1.1.1-1 requires that a sample representative of the tank contents be obtained before it is released. The table specifies the following program: Prior to each batch release, analysis for principal gamma emitters (including all peaks identified by gamma spectroscopy) Once per month, analysis of one batch sample for dissolved and entrained gases (gamma emitters). (See note in Section 2.2.2 below.) { ~ Once per month, analysis of a composite sample' of all releases that month for tritium (H-3) and gross alpha activity. (The composite I sample is required to be representative of the liquids released and sample quantitles of the composite are to be proportional to the quantities of liquid discharged). Once per quarter, analysis of a composite sample of all releases that quarter for Strontium (Sr)-89, Sr-g0, and iron (Fe)-55. f. 2.2.2 CONTINUOUS Releases Fermi 2 TS Table 4.11.1.1.1-1 requires that composite samples be collected from the CWR System, if contaminated. The table specifies the following sample analysis: Once per month, analysis of a composite sample for principal'gamme emitters and for 1-131. Once per month, analysis of a composite sample for H-3 and gross alpha. Once per month, analysis of weekly grab samples (composited) for dissolved and entrained gases (gamma emitters). (See note below.) Once per quarter, analysis for Sr-89, -go and Fe-55. NOTE: Identification of noble gases that are principal gamma emitting radionuclides are included in the gamma spectral analysis performed on all liquid radweste effluents. Therefore, the TS Table 4.11.1.1.1-1 sampling and analysis for noble gases in batch releases (one batch per month) and continuous releases (monthly analysis of weekly grab samples) need not be performed as a separate program. The gamma spectral analysis on each batch release and on the CWR monthly composite meets the intent of this TS requirement. 2.3 Liquid Effluent Monitor Setpoints Technical Specification 3.11.1.1 requires that the concentration of liquid radioactive effluents not exceed the unrestricted area MPC at the discharge point to Lake Erie. Dissolved or entrained noble gases in liquid effluents are limited to a concentration of 2 E-04 uCi/mt, total noble gas activity. TS 3.3.7.11 requires that radiation monitor setpoints be established to alarm and trip prior to exceeding the limits of TS 3.11.1.1.
R: vision 2 Pag 2 2.0-4 To meet this specification, the alarm / trip setpoints for liquid effluent monitors are determined in accordance with the following equation: SP < CL (DF + RR) RR (2-1) where: the setpoint,in uCi/ml, of the monitor measuring the radioactivity SP = concentration in the effluent line prior to dilution. The setpoint represents a value which, if exceeded, would result in concentrations exceeding the MPC in the unrestricted area the effluent concentration limit (TS 3.11.1.1) implementing 10 CFR CL = Part 20.106 (i.e., MPC at discharge point) in uCi/ml, defined in Equation (2-4) the liquid effluent release rate as measured at the radiation monitor RR = location, in volume per unit time, but in the same units as DF, below the dilution water flow as measured prior to the release point (Lake DF = Erie) in volume per unit time At Fermi 2 the available Dilution Water Flow (DF) is constant for a given release, and the waste tank Release Rate (RR) and monitor Setpoint (SP) are set to meet the condition of Equation (2-1) for a given effluent Concentration Limit, CL NOTE: If no dilution is provided, SP < CL Also, when DF is large compared to RR, then (DF + RR)~,DF. 2.3.1 Liquid Radwaste Effluent Line Monitor (D11-N007) Liquid Redweste Effluent Line Monitor D11-N007 provides alarm and automatic termination of releases prior to exceeding MPC. As required by .TS Table 4.11.1.1.1-1 and as discussed in ODCM Section 2.2.1, a sample of the liquid radwaste to be discharged is collected and analyzed by gamma spectroscopy to identify principal gamma emitting radionuclides. From the measured individual radionuclides concentrations, the allowable rolesse rate is determined. e
m ODCM-2.0 R; vision 2 Page 2.0
~~The allowable release rate is inversely proportional to the ratio of the radionuclides concentrations to the MPC values The ratio of the measured l~~
~~concentration to MPC values is referred to as the *MPC fraction~~
~~and is calculated by the eqaation:~~
MPCF. C1 MPC) (2-2) where; MPCF. = fraction of the unrestricted area MPC for a mixture of radionuclides = concentration of each radionuclides i measured in each tank - Ci l prior to release (uCi/ml) = unrestricted area most restrictive MPC for each radionuclides i MPCi from 10 CFR Part 20, Appendix B, Table li, Column 2. For dissolved and entrained noble gases an MPC value of 2E-04 uCl/mi shall be used. Based on the MPCF, the maximum allowable release rate can be calculated by the following equation: MAX RR = DF_ e gp MPCF (2-3) where: MAX RR = maximum acceptable waste tank discharge rate (gal / min) (Monitor #G11-R703) DF = dilution flow rate from the CWR as observed from the Control Room readout (gel / min) (Monitor #N71-R802) SF = 0.5, administrative safety factor to account for vedations in monitor response and flow rates. The S7 value of 0.5 providas for 100% variation caused by statistical fluctuation and/or errors in measurements. Also, this factor provides conservatism, accounting for the presence of radionucl6 des that may not be detected by the monitors (i.e., non-gamma emitters). MPCF = As previously defined by equation 2-2. NOTE: Equation (2-3) is valid only for MPCF >1: if the'MPCF <1, the waste tank concentration meets the limits of 10 CFR Part 20 without dilution, and the waste sample tank may be discharged at the maximum rate. If MAX RR as calculated above is greater than the maximum discharge pump capacity, the pump capacity should be used in establishing the actual
Page 2.0-6 1 Chicase Rate RR f:r tha redweste discharge. Fcr the Weste Sample Tank, the maximum discharge rate is 50 gallons por minute. The actual Release t ) Rate RR is monitored in the Redweste Control Room by 011-R703. The Concentration Limit (CL) of a liquid redweste discharge is the same as the effective MPC for the radionuclides mixture of the discharge. Simply, the ~ CL (or effective MPC) represents the equivalent MPC value for a mixture of radionuclides evaluated collectively. The equation for determining CL is: CL = (2-4) Based on the Release Rate RR and Dilution Flow DF and by substituting Equation (2-4) for CL in Equation (2-1), the alarm setpoint is calculated by the equation: SP, (C1 e SENi)
~~DFa. gkg MPCF e RR (2-5) where:~~
SP = setpoint of the radiation monitor counts per second (cps) = concentration of radionucilde i as measured by gamma f Ci spectroscopy (uCi/ml) = monitor sensitivity for radionuclides i based on calibration j SENi curve (cps /(uCi/ml)) RR = actual release rate of the liquid redweste discharge (gal / min) MPCF = MPC fraction as determined by Equation (2-2) Bkg = background reading of monitor (cps) DF = Dilution flow rate of Circulating Water Decant Line as observed itom Control Room readout (gal / min) monitor #N71-R802. The Cs-137 sensitivity may be used in lieu of the sensitivity values for individual radionuclides. The Cs-137 sensitivity provides a reasonable conservative monitor response correlation for radionuclides of interest In reactor effluents. Coupled with the Safety Factor SF in Equation (2-3). l this simplifying assumption does not invalidate the overall conservatism of the setpoint determination. If no radionuclides are measured by gamma spectroscopy, the alarm setpoint can be established at 2 times the radweste monitor (D11-N007) background. l Prior to conducting eny batch liquid redweste release. Equation (2-3) is used to determine the allowable release rate in accordance with Technical l Specification 3.11.1.1. Equation (2-5) is used to determine the D11-N007 slarm setpoint in accordance with TS 3.3.7.11.
R:;visien 2 Page 2.0-7 2.3.2, ' Circulating Water Reservoir Decent Line Radiation Monitor (D11-N402) Technical Specification 3.3.7.11 requires that the setpoint for the CWR Dec:ent Line Radiation Monitor D11-N402 be established to ensure the radioactive material concentration in the decent line prior to discharge to Lake Erie does not exceed MPC, unrestricted area (10 CFR 20, Appendix B, Table 11. Column 2). The approach for determining the alarm setpoint for the CWR Decent Line Radiation Monitor is the same as presented in Section 2.3.1 for the Liquid Radweste Effluent Line Monitor. Equation (2-1) remains valid, except that, for the CWR Decant Line Monitor, the dilution flow previously assumed for diluting the BATCH liquid redweste effluents is now the release rate. There is no additional dilution prior to discharge to Lake Erie. Thus Equation (2-1) simplifies to: SP < CL f (2-6) i-Substituting Equation (2-4) for CL, the D11-N402 alarm setpoint can be calculated by the equation: SP < C1 MPCF (2-7) where: = concentration of each radionuclides i in the CWR decant line Ci effluent uCi/ml) MPCF = MPC fraction as determined by Equation (2-2) Normally, only during periods of batch liquid radweste discharges will there exist any plant-related radioactive material in the CWR decent line. 2.3.3 Generic, Conservative Alarm Setpoint for D11-N402 i The D11-N402 setpoint could be adjusted for each BATCH release as is done for the liquid radwaste effluent line monitor. Based on the measured levels of radioactive material in a BATCH liquid release, the alarm setpoint for D11-N402 could be calculated using Equation (2-7). However, during r these planned releases, the concentrations will almost always be so low (due to dilution) that the D11-N402 Monitor will not indicate measurable levels. The CWR decent line design flow is 10,000 ppm; and the maximum liquid redweste release rate is 50 ppm, providing a 200:1 dilution. The the rangt of 10-gal conegntration of BATCH liquid releases is typica radioactive mate to 10- UCl/ml. With a nominal 200:1 dilution, the CWR ogant line myttor would monitor diluted activity in the range of 5 x 10- to 5 x 10- uCl/ml. D11-N402 Monitor response at these levels would be 10 to 100 cpm, depending on the particular radionuclides mixture and corresponding instrument response. These response levels are p less than the monitor background levels. In lieu of routinely adjusting the D11-N402 storm setpoint, a generic, conservative alarm setpoint can be established. The Fermi 2 UFSAR, l 1
ODCM-2.0 Revision 2 Page 2.0-8 Section 11.2, Table 11.2-g, presents the estimated releases of radioactive liquid offluents. Using Equation (2-4), the distribution of the estimated releases corresponds to a Concentration Limit CL (or effective MPC) of - 3 E-06 uCl/ml. Using the Cs-137 sensitivity of 2.05 E + 08 cpm /uCi/mi and CL = 3 E-06 uCi/ml, the corresponding D11-N402 alarm setpoint is 615 cpm above monitor background. Refer to Appendix A for details on the determination of CL F.3.4 Alarm Setpoint for GSW and RHR System Radiation Monitors l Levels of radioactive material detectable above background at Radiation Monitor D11-N008 would be one of the first indicators of contamination nf the General Service Water (GSW) System and the CWR. Likewise, for the Residual Heat.'temoval (RHR) System, the D11-N401 A and B Monitors would be one of the first indicators of contamination and subsequent contamination of the CWR. Therefore, to provide early indication and assure prompt ettention, the alarm setpoints for these monitors should be established as close to background as possible without incurring a spurious alarm due to background fluctuations. This level is typically around three times background. If the GSW System or AHR System becomes contaminated, it may become j necessary to raise the radiation monitor setpoints. The alarm se! points should be re-evaluated to provide the CR operator a timely indication'of further increasing activity levels in the GSW or RHR System whho Jt spurious alarms. The method for this re-evaluation is the same as described above - the alarm setpoint established at three times its current reading. No regulatory limits apply for establishing a maximum value for these alarm setpoints since these monitors are located on plant systems and do not monitor final rolesse points to the environment. However, as a practical matter, upper limits on the alarm setpoints can be evaluated using the methods of ODCM Section 2.3.1 based on the actual system flows, dilution and release paths in effect at the time. 2.3.5 Alarm Response - Evaluating Actual Release Conditions Normally, liquid release rates are controlled and alarm setpoints are established to ensure that the release does not exceed the concentration limits of TS 3.11.1.1 (i.e.,10 CFR 20 MPCs at the discharge to Lake Er!e). However, if either Monitor D11-N007 or D11-N402 alarms during a liquid release, it becomes necessary to re-evaluate the release conditions to determine compliance with TS 3.11.1.1. Following an alarm, the actual rolesse conditions should be determined. Radioactive material concentrat;ons should be evaluated by sempling the effluent stream or resampling the weste tank. Discharge flow and dilution water flow should l be redetermined. The following equation may be used for the evaluation: I C1 3 RR, ,g 1 i (uPCy or inR (2-8) where:
L Pag) 2.0-g ' Ci = measured co'ncentration of radionuclides i in the effluent stream (uCl/ml) MPCl = the MPC value for red'ionuclide i from.10 CFR 20, Appendix B, Table 11, Column 2 (uCi/ml),2 E-04 uCi/mi for dissolved or l entrained noble gases RR =- actual release rate of the liquid effluent at the time of the i alarm, gpm I J DF = actual dilution circulating water flow at the time of the release alarm, gpm o NOTE: For alarm on D11-N402 (CWR decent line), the Release Rate RR is the Dilution Water Flow DF and the equation simplifies to (C /MPC)) St. ,f i 2.3.6 Liquid Radweste Monitor Setpoint Determination with Contaminated Circulating Water Reservoir in the event the CWR is determined to contain radioactive material, the. effective dilution capacity of the CWR is reduced as a function of the' MPCF. To determine the available dilution flow capacity the MPCF for the CWR is determined using equation (2-2). The MPCF of the CWR is used to determine the available dilution flow as follows: CWR Dilution Flow = CWR Decant Flow Rate (OPM) * (1-CWR MPCF) (2-g) The resulting dilution flow rate is substituted in equation (2-3) to determine the maximum allowable release rate for discharges from the radwaste. system. Substituting the available CWR dilution flow from equation (2-g), the Liquid Radweste Monitor maximum release rate can be determined using equation (2-3). Once the available dilution flow and maximum allowable release rate have .been determined the redweste monitor setpoint can be determined using - equation (2-5). 2.4 Contaminated GSW or RHR System - Quantifying and Controlling Releases The GSW Radiation Monitor (D11-N008) provides an indication of contamination of l this system. The Monitors D11-N401 A and B perforrn this function for the RHR System. Also, the CWR Decant Line Radiation Monitor monitors all liquid releases from the p.lant and would record any release to Lake Erie from either of these systems if contaminated. As discussed in ODCM Section 2.2.2, sampling and analysis of the CWit System is required only if this system is contaminated, as would be indicated by D11-N402 or D11-N008. Nonetheless, periodic samples are collected from the CWR System to verify absence of contamination. Although not required by TS, periodic sampling and analysis of the RHR System is also performed since it also is a potential source of contamination of the CWR and subsequent releases to Lake Erie. If contamination is found, further rolesses from the applicable system (GSW or RHR) via the CWR decent line must be evaluated and controlled to ensure
ODCM-2.0 C: vision 2. P ge 2.0-10 that releases are maintained ALARA. The following actions will be considered for controlling releases. Sampling frequency of the applicable source (GSW or RHR System) and the CWR will be increased until the source of the contamination is found and' controlled. This frequency may be relaxed after the source of contamination has been identified and isolated. Gamma spectral analysis will be performed on each sample. The measured radionuclides concentrations from the gamma spectral analysis will be compared with MPC (Equation 2-2) :o ensure releases are within the limits of TS 3.11.1.1. Based on the measured concentrations, the setpoint for the CWR Decent Line Radiation Monitor (011-N402) will be determined as specified in Section 2.3.2. If the calculated setpoint based on the measured distribution is greater than the current setpoint (see ODCM Section 2.3.3) no adjustment to the setpoint is required. Samples will be composited in accordance with TS Table 4.11.1.1.1-1 for monthly analysis for H-3 and gross alpha and for quarterly analysis for Sr-89, g0 and Fe-55. Each sample will be considered representative of the releases that have occurred since the previout sample. For escli sample (and corresponding release period). the volume of liquid released to the lake will be determined based on the measured CWR decent line cumulative flow. From the sample analysis and the calculated volume released, the total radioactive meterial released will be determined and considered representative of the release period. Cumulative doses will be determined in accordance with ODCM Section 2.5. 2.5 Liquid Effluent Dose Calculation - 10 CFR 50 The parameters of the liquid release (or estimated parameters. for a pre-release calculation) may be used to calculate the potential dose to the public from the release (or planned release). The dose calculation provides a conservative method for estimating the impact of radioactive effluents released by Fermi 2 and for comparing that impact against limits set by the NRC in the Fermi 2 TS. The limits in the Fermi 2 TS are specified as quarterly and calender year limits. This assures that the average over the year is kept as low as reasonably achievable. 2.5.1 MEMBER OF THE PUBLIC Dose - Liquid Effluents Technical Specification 3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBUC from radioactive materials in liquid effluents from Fermi 2 to: during any calender quarter; $ 1.5 mrom to total body i i 5.0 mrom to any organ f.
R:visi:n 2 Page 2.0-11 during any calender year; $ 3.0 mrom to total body $ 10.0 mrom to any organ The calculation of the potential doses to MEMBERS OF THE PUBUC is a ( function of the radioactive material releases to the lake, the subsequent transport and dilution in the exposure pathways, and the resultant individual uptake. At Fermi 2, pre-operational evaluation of radiation exposure l pathways Indicated that doses from consumption of fish from Lake Erie provided the most conservative estimate of doses from releases of radioactive liquids. However, with the proximity of the water intake for the City of Monroe. It must be assumed that individuals will (;onsume drinking water as well as fish that might contain radioactivity from discharges into Lake Erie. Study of the currents in Lake Erie indicates that the current in the Lagoona Beach embayment carries liquid effluents from Fermi 2 north along the coast part of the time and south along the coast part of the time. When the current flows north, liquid affluents are carried away from the Monroe Water intake, so only the fish consumption exposure pathway must be considered. When the current flows south, toward the Monroe Water intake, both fish consumption and drinking water consumption exposure pathways must be considered. To ensure conservatism in the dose modeling, the combined fish and drinking water pathway is used for j evaluating the maximurn hypothetical dose to a MEMBER OF THE PUBLIC I-from liquid radioactive effluents. The following calculational methods may be used for determining the dose or dose commitment due to the liquid radioactive effluents from Fermi 2: De, = 1.67 E-02
VOL e (C) Ago; DF

Z (2-10) where:
dose or dose commitment to orgen o or total body (mrem) Do = site-specific ingestion dose commitment factor to the total Ao = i body or any organ o for radionuclides I (mrom/hr per uCi/m!) average concentration of radionuclida i in undiluted liquid C; = effluent representative of the volume VOL (uCl/ml) total volume of liquid offiuent released (gal) VOL = average dilution water flow (CWR decant line) during DF = release period (gal / min) 5, near field dilution factor Z = (Derived from Regulatory Guide 1.109, Rev 0) 1 hr/60 min 1.67 E-02 = l
R3 vision 2 P:ge 2.0-12 The site-specific engestion dose / dose commitment fact:rs (Ago) r:pras nts i a composita dose factor for the fish and drinking water pathway. The site-specific dose factor is based on the NRC's generic maximum individual consumption rates. Values of Ago are presented in Table 2-1. They were derived in accordance with guidance of NUREG-0133 from the following equation: Ae e 1.14 E + 95 (Uw / Dw + Up e BF ) DF) 6 6 (2-11) where: l 21 kg/yr adult fish consumption l Up = 730 Nters/yr adult water consumption -Uw = 15.4, additional dilution from the near field to the water DW = intake for the City of Monroe (Not dilution factor of 77 from discharge point to drinking water intake, Fermi 2 UFSAR, Chapter 11, Table 11.2-11) Bioaccumulation factor for radionuclides i in fish from BFj = Table 2-2 (pCi/kg per pCi/ liter) dose conversion factor for nuclide i for adults in organ o DF) = from Table E-11 of Regulatory Guide 1.109 (mrom/pCl) 8 106 (pCi/uCl)
~~10 (ml/kg) 1.14 E + 05~~
= 8760 (hr/yr) The radionuclides included in the periodic dose assessment required by TS 3.11.1.2 are those identified by gamma spectral analysis of the liquid waste samples collected and analyzed per the requirements of TS Table 4.11.1.1.1-1. In keeping with the NUREG-0133 guidance, the aduft age group represents the maximum exposed individual age group. Evaluation of doses for other age groups is not required for demonstrating compliance with the dose criteria of TS 3.11.1.2. The dose analysis for radionuclides requiring radiochemical analysis will be performed after receipt of results of the analysis of the composite samples, in keeping with the required analytical frequencies of TS Table 4.11.1.1.1-1, tritium dose analyses will be petformed at least monthly: Sr-89, Sr-90 and Fe-55 dose analyses will be performed at least quarterly. 2.5.2 Simplified Liquid Effluent Dose Calculation in lieu of the individual radionuclides dose assessment presented in Section 2.5.1, the following simplified dose calculation may be used for demonstra, ting compliance with the dose limits of TS 3.11.1.2. (Refer to Appendix Is for the derivation of this simplified method.) f
P:ge 2.0-13 Total Body Ci Deb = 9.99 E + 03
~~VOL DF e Z (2-12).~~
~~Maximum Organ e Ci Dmex = 1.18 E + 04~~
~~VOL DF e 2 (2-13) where:~~
C1 = everage concentration of radionuclides I in undiluted liquid effluent representative of the volume VOL (uCi/ml) VOL = volume of undiluted liquid offluent released (gal) DF = average dilution water flow (CWR decent line) during release period (gal / min) Z = 5, near field dilution factor (derived from Regulatory Guide 1.109, F;ev 0) Dtb = conservatively evaluated total body dose (mrom) Dmax = conservatively evaluated maximum organ dose (mrom) g.69 E + 03 = 0.0167 (hr/ min)
~~5.80 E + 05 (mrom/hr per uCl/ml, Cs-134 total body dose factor from Table 2.0-1) 1.18 E + 04~~
~~= 0.0167 (hr/ min)~~
7.09 E + 05 (mrom/hr per uCi/ml, Cs-134 liver dose factor from Table 2.0-1) 2.5.3 Contaminated CWR System - Dosa Calculation if the CWR System becomes contaminated, releases via the CWR System to Lake Erie must be included in the evaluation of the cumulative dose to a MEMBER OF THE PUBLIC as required by TS 3.11.1.2. ODCM Section 2.4 described the methods for quantifying and controlling releases from the CWR System.
For calculating the dose to a MEMBER OF THE PUBLIC, Equation (2-10) remains applicable for releases fro:n the OSW System with the following assumptions: DF, Dalution Flow, is set equal to the average CWR decent line flow rate over the release period. Cg. Radionuclides Concentration,is determined as specified in ODCM Section 2.4.
P:ge 2.0-14 VOL Valum) Releas:d. is s;t equal to the total vslume of the discharges to Lake Erie via the CWR decent line as specified in Section 2 4. .2.6 Liquid Effluent Dose Projections 10 CFR 50.36s requires licensees to maintain and operate the Radweste System to ensure releases are maintained ALARA. This requirement is implemented through TS 3.11.1.3. This TS requires that the Liquid Radioactive Weste Processing System be used to reduce the radioactive material levels in the liquid waste prior to release when the projected dose in any 31 day period would exceed: 0.06 mrom to the total body, or 0.2 mrom to any organ When the projected doses exceed either of the above limits, the waste must be processed by the Liquid Redweste System prior to release. This dose criteria for processing is established at one forty eighth of the design objective inte (3 mrom/yr, total body or 10 mrom/yr any organ) in any 31 day period. The applicable Liquid Waste Processing System for maintaining radioactive material releases ALARA is the lon Exchange System as delineated in Figure 2-1. Alternately, the Waste Evaporator (presented in the Fermi 2 UFSAR, Section 11.2) can be used to meet the NRC ALARA design requirements it may be used in conjunction with or in lieu of the ton Exchange System to meet the waste processing requirements of TS 3.11.1.3. Each BATCH release of liquid radweste is evaluated to ensure that cumulative doses are maintained ALARA. In keeping with the requirements of TS 3.11.1.3, dose projections are made at least once per 31 days to evaluate the need for cdditional radwaste processing to ensure future releases are maintained ALARA. The following equations may be used for the dose projection calculation: Dtbp = Dtb (31 / d) j Dmaxp = Dmax (31/ d) (2-15) where: Dtbp = the total body dose projection for current 31 day period (mrom) Db = the comulative total body dose to date for current calendar quarter t including release under consideration as determined by equation (2-10) or (2-12) (mrom) Dmaxp = the maximum organ dose projection for current 31 day period (mrom) Dmax = the maximum organ dose to date for current calendar quarter including release under consideration as determined by Equation (2-10) or (2-13) (mrom) 1
Pag) 2.0-15 l d = the numb:r of days 12 date in currcnt cal:ndar quarter i = the number of days in projection 31 END OF SECTION 2.0 e 1 0 0 (
ODCM-2.0 Revisi:n 2 Ptge 2.0-15 TABLE 2.0-1 Fermi 2 Site Specific Liquid Ingestion Dose Commitment Factors A o (mrom/hr per uCi/ml) i C!uclide Bone Uver T Body Thyroid Kidney Lung GI-LLI 7.94E-1 7.94E-1 7.94E-1 7.94E-1 7.94E-1 7.94E-1 H-3 C-14 3.13E+4 S.26E+3 6.26E+3 6.26E+3 6.26E+3 6.26E+3 6.26E+3 - Nn-24 4.16E+2 4.16E+2 4.16E+2 4.16E+2 4.16E+2 4.16E+2 4.16E+2 1.56E+ 5 P-32 1.39E+6 8.62E+4 5.36E+4 1.29E+0 7.70E-1 2.84E-1 1.71E+0 3.24E+2 Cr-61 1.35E+4 1.31E+3 4.40E+3 8.40E+2 Mn-54 3.53E+3 1.41E+2 1.11E+2 1.96E+1 Mn-56 2.59E+2 2.67E+2 FO-55 6.73E+2 4.65E+2 1.08E+2 6.98E+2 8.32E+3 Fo-59 1.06E+3 2.50E+3 9.57E+2 5.55E+2 2.19E+1 3.64E+1 Co-57 1.89E+3 9.32E+1 2.09E+2 Cc-58 5.03E+3 2.68E+2 5.90E+2 Co-60 4.60E+2 Ni-53 3.18E+4 2.21E+3 1.07E+3 4.26E+2 Ni-65 1.29E+2 1.68E+1 7.66E+0 8.88E+2 2.63E+1 1.04E+1 4.89E+0 Cu-64 4.65E+4 4.94E+4 Zn-65 2.32 E+4 7.38E +4 3.34E+4 1.42E+ 1 6.14E+1 Zn-69 4.94E + 1 9.44E+1 6.57E+0 2.62 E+3 2.28E+3 Br-82 5.85E+1 4.06E+ 1 Br-83 4.13E-4 5.27E+1 Cr-84 1.01E-15 2.16E+0 Br-85 1.99E+4 1.01E+5 4.71E+4 Rb-86 A.01E-9 2.90E+2 1.54E+2 [ Cb-88 1.12E-11 1.92E+2 1.35E+2 Rb-89 3.81 E+3 6.83E+2 Sr-89 2.38E+4 1.69E+4 1.44E+5 Sr-90 5.85E+5 2.09E+3 1.77E+1 Sr-91 4.38E+2 3.29E+3 7.18E+0 Sr-92 1.66E+2 6.66E+3 1.68E-2 Y-90 6.28E-1 1.74E-2 2.30E-4 Y-91m 5.93E-3 5.06E+3 2.46E-1 Y-91 9.20E+0 9.66E+2 1.61E-3 Y-92 5.51E-2 5.55E+3 4.83E-3 Y-93 1.75E-1 4.11E+2 l 2.04E-1 Zr-95 4.04E-1 1.30E-1 8.78E-2 1.40E+3 6.81E-3 Zr-97 2.24E -2 4.51E-3 2.06E-3 1.51E+6 2.46E+2 Nb-95 4.47E+2 2.49E+2 1.34E+2 3.50E+3 1.11E+0 Nb-97 3.75E+0 9.48E-1 3.46E-1 2.93E+2 2.86E+2 1.26E+2 2.41E+1 Mo-99 4.38E-1 1.41E-2 1.71E+1 Tc-99m 1.02E-2 2.88E-2 3.67E-1 2.72E-1 7.73E-3 4.54E-14 Tc-101 1.05E-2 1.51E-2 1.48E-1
1 CDCM-2.0 C:visi n 2 Pa93 2.0-16 TABLE 2.0-1 Fermi 2 Site Specific Liquid Ingestion Dose Commitment Factors A6e (mrom/hr per uCi/ml) Nuclide Bone Liver T Body Thyroid Kidney Lung GI-LLI 6.34E+2 2.07E+1 2.34E+0 Cu-103 5.43E+0 2.76E+2 5.84E+0 1.78E-1 Ru-105 4.52E-1 5.22E+3 1.56E+2 1.02E+1 Ru-106 8.07E+ 1 Rh-103m Rh-106 6.59E+2 3.17E+0 A9-170m 1.75E+0 1.61E+0 9.59E-1 1.70E+1 6.20E+2 Sb-124 2.18E+1 4.13E-1 8.66E+0 5.29E-2 1.08E+1 1.54E+2 .Sb-125 1.40E+1 1.56E-1 3.32E+0 1.42E-2 1.03E+4 To-125m 2.58E+3 9.35E+2 3.46E+2 7.76E+2 1.05E+4 2.19E+4 To-127m 6.52E+3 2.33E+3 7.94E+2 1.67E+3 2.65E+4 8.36E+3 Tc-127 1.06E+2 3.80E+1 2.29E+1 7.85E+1 4.31E+2 5.58E+4 To-129m 1.11E+4 4.13E+3 1.75E+3 3.80E+3 4.62E+4 2.28E+1 Ta-129 3.02E +1 1.14E+1 7.37E+0 2.32E+1 1.27E+2 8.09E+4 Tc-131m 1.67E+3 8.15E+2 6.79E+2 1.29E+3 8.25E+3 2.69E+0 To-131 1.90E+1 7.93E+0 5.99E+0 1.56E+1 8.31E+ 1 7.42E+4 To-132 2.43E+3 1.57E+3 1.47E+3 1.73E+3 1.51E+4 7.93E+ 1 I-130 3.12E + 1 9.21E+1 3.64E+1 7.81E+3 1.44E+2 f. 6.49E+1 4-131 1.72E+2 2.46E+2 1.41 E+2 8.06E+4 4.21E+2 4.21E+0 1-132 8.39E+0 2.24E+ 1 7.85E+0 7.85E+2 3.57E+1 9.17E+1 1-133 5.87E+ 1 1.02E+2 3.11E+1 1.50E+4 1.78E+2 1.04E-2 1-134 4.38E +0 1.19E+ 1 4.26E+0 2.06E+2 1.89E+1 5.41E+1 1-135 1.83E+1 4.79E+1 1.77E+1 3.16E+3 7.68E+1 2.30E+5 7.62E+4 1.24E+4 Cn-134 2.98E+5 7.09E+5 5.80E+5 6.85E+4 9.40E+3 1.40E+4 Cs-136 3.12E+4 1.23E+5 8.87E+4 1.77E+5 5.90E+4 1.01E+4 Cs-137 3.82E+5 5.22E+5' 3.42E+5 3.84E+2 3.79E+ 1 2.23E-3 Cs-138 2.65E+2 5.22E+2 2.59E+2 9.68E-4 5.87E-4 2.58E+0 Ba-139 1.45E+0 .1.04E-3 4.25E-2 1.30E-1 2.19E-1 6.26E+2 Ba-140 3.04E+2 3.82E-1 1.99E+1 4.96E-4 3.03E 3.33E-10 Be-141 7.06E-1 5.33E-4 2.38E-2 2.77E-4 1.86E-4 - 4.49E-19 Ba-142 3.19E-1 3.28E-4 2.01E-2 6.04E+3 l La-140 1.63E-1 8.22E-2 2.17E-2 2.77E+1 l La-142 8.35E-3 3.80E-3 9.46E-4 1.89E+2 2.29E-2 Co-141 7.30E-2 4.94E-2 5.60E-3 3.56E+2 4.19E-3 Co-143 1.29E-2 9.51E+0 1.05E-3 1.29E+3 9.44E-1 l Co-144 3.81E+0 1.59E+0 2.04E-1 2.63E+3 1.39E-1 Pr-143 6.00E-1 2.41 E-1 2.98E-2 2.83E-10 ) 4.60E-4 Pr-144 1.96E-3 8.16E-4 9.98E-5 2.28E+3 2.77E-1 Nd-147 4.10E-1 4.74E-1 2.84E-2 8.12E+4 W-187 2.96E+2 2.48E+2 8.66E+1 7.04E+2 1.07E-2 Np-239 3.49E-2 3.43E-3 1.89E-3 l
Revisi:n 2 ) . Pega 2.0-17 ~ TABLE 2.0-2 Bioaccumulation Factors (SFi) ] (pC4/kg per pCl/ liter)* Element Freshwater Fish H 9.0E-01 C 4.6E+03 Na 1.0E+02 P-3.0E+03 Cr 2.0E+02 Mn 4.9E+02 Fe 1.0E+02 Co 5.0E+01 1 i Ni 1.0E+02 Cu 5.0E+01 Zn 2.0E+03 Br 4.2E+02 Rb 2.0E+03 Sr 3.0E+01 Y 2.5E+01 Zr 3.3E+00 Nb 3.0E+04 Mo 1.0E+01 Tc 1.5E+01 Ru 1.0E+01 Rh 1.0E+01 Ag 2.3E+00 Sb 1.0E+00 Te 4.0E+02 1 1.5E+01 Cs 2.JE+03 Ba 4.0E+00 Le 2.5E+01 Ce 1.0E+00 Pr 2.5E+01 Nd 2.5E+01 W 1.2E+03 Np 1.0E+01 i
~~Values in this table sie taken from Regulatory Guide 1.109 except for phosphorus, which is adapted from NUREG/CR-1336, and silver and antimony, which are taken from UCRL 60564 Rev 1, October 1972.~~
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Nuclear Production - Fermi 2 00CM-3.0 Off:lt3 Do03 Calculation Manual Revi:12n 2 Page 3.0-1 GASEOUS EFFLUENTS 3.0 GASEOUS EFFLUENTS 3.1 Radiation Monitoring instrumentation and Controls 3.1.1 Sffluent Monitoring - Ventilation System Releases The gaseous effluent monitoring instrumentation required at Fermi 2 for controlling and monitoring radioactive affluents are specified in TS 3.3.7.12. The monitoring of each identified gaseous effluent release poirit must include the following: Noble Oas Activity Monitor iodine Sampler (sample cartridge containing charcoal or silver roollte) Particulate Sampler (filter paper) Sampler Flow Rate Monitor Meeting these requirements, a total of seven Eberline SPING Monitoring Systems are installed on the six gaseous release points (Onsite Storage Facility, Service Building. Radweste Building. Turbine Building, Reactor Building Exhaust Plenum, end Standby Oss Treatment System Division 1 and Division 2). The SPING Monitor cutputs are recorded electronically in the CT-28 Control Terminal in the Main Control Room. In general, a reading exceeding the High alarm setpoint of the SPING Monitors causes an alarm in the Control Room. Fermi 2 TS Table 3.3.7.12-1 identifies only the alarm function of the ReactM Building Exhaust Plenum Effluent Monitor, the Standby Oss Treatment System Monitors, and the Onsite Storage Facility. 3.1.2 Main Condenser Offgas Monitoring TS Table 3.3.7.12-1 specifies monitoring requirements for the Offges System at the 2.2 minute delay line. The following monitors are requireo: Hydrogen Monitor - used to ensure the hydrogen concentration in the Offges Treatment System is maintained less than 4% by volume as required by TS 3.11.2.6. Noble Oss Activity Monitor - used to ensure the gross activity release rate is maintained within 340 millicuries per second after 30 minute decay as required by TS 3.11.2.7. ARMS - DEFORMATION SERVICES Date approved: Release authorized by:_ Changs numbers incorporated: 1.CR 88-032-ODM DSN Rev 2 Date DTC TMPLAN File 1715.02 Recipient 1
ODCM-3.0 Revision 2 Page 3.0-2 I These two monitors perform safety functions. The Hydrogen Monitor monitors the potential explosive mixtures in the Offges System. The Noble Oas Monitor monitors the release rate from the main condenser ensuring doses at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment bypassing the Offges Treatment System. 3.1.3 Reactor Building Ventilation Monitors (Gulf Atomic) The Outf Atomic Monitors (D11-N408 and 410) on the Reactor Building Ventilation System provide on high radiation levels (above alarm setpoint) initiation of SGTS, isolation of drywell vent / purge, isolation of the RB and Control Center Ventilation Systems and initiation of Control Center recirculation modo ventilation. These monitors and functions are not required by Fermi 2 TS but are important in controlling containment venting / purging. 3.2 Samplinp and Analysis of Gaseous Effluents The program for sampling and analysis of gaseous waste is prescribed in Fermi 2 TS Table 4.11.2.1.2-1. ' This table distinguishes two types of gaseous releases: (1) containment PURGE, treated as BATCH releases, and (2) discharges from the Reactor Building Exhaust Plenum (including Standby Gas Treatment System (SGTS) when operating), and other building ventilation exhausts, treated as CONTINUOUS releases. 3.2.1 Containment PURGE TS Table 4.11.2.1.2-1 requires that a grab sample be collected and analyzed before each containment drywell PURGE. Sampling and analysis are required within eight hours before starting a PURGE. TS Table 4.11.2.1.2-1 Footnote I and TS 4.11.2.8.3 also require that if the PURGING or VENTING is through the Reactor Building Vent, rather than through SGTS, additional sample and analyses are required every twelve hours throughout the release period. Analysis must include principal gamma emitters and tritium prior to venting and purging. For a planned containment PURGE, the results of the sample and analysis are used tu establish the acceptable release rate and radiation monitor alarm setpoint in accordance with ODCM Section 3.3. This evaluation is necessary to ensure compliance with the dose rate limits of TS 3.11.2.1. The periodic samples collected throughout the PURGENENT period are used to ensure that release conditions over an extended period are maintained within TS limits. 3.2.2 Ventilation System Releases TS Table 4.11.2.1.2-1 requires e,ontinuous samples of releases from the RB Exhaust Flenum Standby Oss Treatment System, Redweste Building, Turbine Building, Service Building, and Onsite Storage Facillity. The table specifies the following program: Once per wealt analysis of an adsorbent sample of I-131 and 1-133, plus analysis of a particulate sample for principal gamma emitters.
Revision 2 Page 3.0-3 Once per month, analysis of a composite particulate sample of all releases (by release point) that month for gross sipha activity. Once per quarter, analysis of a composite particulate sample of all releases that quarter for Sr-89 and Sr-90. Once per month, analysis of a grab sample for principal gamma emitters (noble gases and tritium). TS Table 4.11.2.1.2-1 also requires continuous monitoring for noble gases. This requirement is met by the SPING Monitors On each of the plant gaseous release points. The TS require more frequent sampling and analysis following reactor startup, shutdown, or change in thermal power exceeding 15% within one hour. The TS allow exceptions to this increased sampling schedule if the Primary Coolant Dose Equivalent 1-131 has not increased more than a factor of three and the Noble Gas (i.e., Offges) Monitor reading has not increased more than a factor of three. Grab samples of the Fuel Pool Ventilation Exhaust are required for tritium analysis once per seven days whenever spent fuel is in the Spent Fuel Pool. Also, grab samples for tritium are required when either the reactor well or the dryer separator pool is filled. These samples are taken at the Reactor Building Exhaust Plenum and Standby Gas Treatment System (SGTS) when operating. 3.3 Gaseous Effluent Monitor Setpoint Determination 3.3.1 Ventilation System Monitors Per the requirements of TS 3.3.7.12, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of TS 3.11.2.1. This TS limits releases to a dose rate at the SITE BOUNDARY of 500 mrom/ year to the total body or 3000 mrom/ year to the skin. From a grab sample analysis of the applicable release (i.e., grab sample of the Drywell or Ventilation System release), the radiation monitoring alarm setpoints may be established by the following calculational method. The measured radionuclides concentrations and release rate are used to calculate the fraction'of the allowable release rate, limited by TS 3.11.2.1, by the equation: FRAC s 1.67 E + 01
X/O

VF * [(C1

Ka) 500 (3-1)
~~FRAC = 1.67 E + 01~~
X/O

VF * [(C1

TL1 + 1.1 Mil) 0 (3-2)
Where: FRAC = fraction of the allowable rolesse rate bas 3d on the identified radionuclides concentrations and the release flow rate
ODCM-3.0 Rovisisn 2 Page 3.0-4 X/O = annual average meterological dispersion to the controlling site boy)dery location from Table 3.0-4 (sec/m VF = Ventilation System flow rate for the applicable release point and monitor (liters / minute) Ci = concentration of noble gas radionuclides i as dotarmined by gamma spectral analysis of grab sample (uCi/cc) = total body dose conversion factorf,or noble gas Ki radionuclides 1 (mrom/yr per uCi/m from Table 3.0-3) Li = beta skin dose conversion factor fgr noble gas radionuclides I (mrom/yr por uCl/m, from Table 3.0-3) Mi = gamma air dose conversion factogfor noble gas radionuclides I (mrad /yr per uCi/m, from Table 3.0-3) 1.1 = mrom skin dose per mrad gamma air dose (mrom/ mrad) 500 = total body dose rate limit (mrom/yr) 3000 = skin dose rate limit (mrom/yr) 1.67 E + 01 = 1 E + 03 (cc/ liter) * (1/60) (min /sec) Based on the more limiting ('i.e., higher) value of FRAC as determined above, the alarm setpoints for the applicable monitors may be calculated by the equation: SP =(AFe Cl) + Skg FRAC (3-3) Where: SP = alarm setpoint corresponding to the maximum allowable l-release rate (uCl/cc) Bkg = background of the monitor (uCi/cc) AF = administrative allocation factor (Table 3.0-2) for the - specific monitor and type release, which corresponds to the fraction of the total allowable release rate that is administratively allocated to the individual release points. concentration of Noble Oss Radionuclides I as determined Ci = by gamma spectral analysis of grab sample (uCl/cc) l l
@2 Page 3.0-5 Th3 Allocati:n Fcctor (AF) is an cdministrativ3 ctntral imp 3 sed to cnsure that combined releases from all release points at Fermi 2 will not exceed the regulatory limits on release rate from the site (i.e., the release rate limits of Technical Specification 3.11.2.1). From the Formt 2 design-evaluation of gaseous effluents presented in the UFSAR Section 11.3, representative values have been determined for AF. These values are presented in Table 3.0-2. These values may be changed in the future as warranted by operational experience, provided the site releases comply with TS 3.11.2.1. When combined with the Noble Oas Monitor calibration l constant, the monitor sensitivity for Xe-133 may be used in lieu of the sensitivity values for the individual radionuclides. Because of its lower gamma energy and corresponding monitor response, the Xe-133 sensitivity provides a conservative value for alarm setpoint determination. I 3.3.2 Conservative, Generic Alarm Setpoints A conservative alarm setpoint can be established,in lieu of the individual l radionuclides evaluation (described above) based on the grab sample analysis. This approach eliminates the need to adjust the setpoint periodically to reflect minor changes in radionuclides distribution or release flow rate. The alarm setpoint may be conservatively determined based on the UFSAR design radionuclides distribution values as summarized in Table 3.0-1. For the radionuclides distribution given in UFSAR Table 11.3-5, the estimated total body dose rate is higher than the estimated skin dose rate. Therefore, the more restrictive setpoint is based on the total body dose rate limit and is calculated with Equations (3-1) and (3-3). The calculated setpoints are presented in Table 3.0-2. 3.3.3 Gaseous Effluent Alarm Response - Evaluating Actual Release Condition The monitor alarm setpoint is used as the primary method for ensuring and demonstrating compliance with the release rate limits of TS 3.11.2.1. Not exceeding alarm setpoints constitutes a demonstration that release rates have been maintained within the TS limits. When en effluent Noble Gas Monitor exceeds the alarm setpoint, en evaluation of compliance with the release rate limits must be performed using actual release conditions. This evaluation requires collecting a sample of the effluent to establish actual radionuclides concentrations and permit evaluating the monitor response. The following equations may be used for evaluating compliance with the release rate limit of TS 3.11.2.1a: Dtb = 1.67 E + 01 e X/O
VF *
~~(Kg~~
~~Cg)~~
(3-4) Ds = 1.67 E + 01 e X/O e VF * (L.; + 1.1 M l e Cg) i (3-5) Where: Ob = total body dose rate (mrom/yr) t l
-- eceua w:r-R: vision 2 Page 3.0-6 Ds = skin dose rete (mrem' /yr) = atmospheric dispersion to the cy) rolling l X/O SITE BOUNDARY location (sec/m VF = Ventilation System rolesse rate (liters / min) I = concentration of radionuclides i as measured in the grab Ci sample or as correlated from the SPING Noble Gas Monitor reading (uCl/cc) = total body dose conversion factorf,or noble gas Ki radionuclides I (mrom/yr por uCl/m from Table 3.0-3) La = beta skin dose conversion factor fgr noble gas radionuclides I (mrom/yr per uCi/m, from Table 3.0-3) Mi = gamma air dose conversion factogfor noble gas radionuclides I (mrad /yr per uCl/m, from Table 3.0-3) 1.1 = mrom skin dose per mrad gamma air dose (mrom/ mrad) 1 E + 03 (cc/ liter) * (1/60) (min /sec) 1.67 E + 01 = .j 3.4 Containment Drywell VENTING and PURGING \\ 3.4.1 Release Rate Evaluation For a drywell VENTING or PURGING, an evaluation of acceptable release rate . should be performed prior to the release. Based on the measured noble gas concentration in the grab sample collected per the requirements of TS Table 4.11.2.1.2-1, the allowable release rate can be calculated by the following equation: RRtb = 500e AF 1.67 E + 01 e X/O e { (Kg
~~Cg)~~
(3-6) or RRs = 3000e AF 1.67 E + 01 e X/O e { ([Li + 1.1 M l e Cg) i (3-7) 1 Where: = allowable release rate so as not to exceed a dose rate RRtb of 500 mrom/yr, total body (liters / minute) RRs = silowa.ble release rate so as not to exceed a dose rate of 3000 mrom/yr, skin (liters / minute) AF = allocation factor for the applicable release point from Table 3.0-2 (default value is 0.5 for Reactor Building Exhaust Plenum) E
Revision 2 Pap 2 3.0-7 500 = total body dose rate limit (mrom/yr) 3000 = skin dose rate limit (mrom/yr) The lesser value (RRtb or RRs) as calculated above should be used for establishing the allowable release rate for the drywell PURGING or VENTING. 3.4.2 Alarm Setpoint Evaluation For a containment drywell VENTING or PURGING, a re-evaluation of the. alarm setpoint is needed to ensure compliance with the requirements of TS 3.3.7.12. For the identified release path (RB Exhaust Plenum or SGTS) and associated effluent Radiation Monitor, the alarm setpoint should be calculated using Equations (3-1), (3-2) and (3-3). In Equations (3-1) and (3-2), the value of the Ventilation Flow VF should be established at the total release flow rate, including the contribution from the PURGE or VENT. If the calculated alarm setpoint is greater than the current setpoint, no adjustments are necessary, 3.5 Quantifying Releases - Noble Gases The determination of doses in the environment from releases is dependent on the mixture of the radioactive material. Also, NRC Regulatory Guide 1.21 requires reporting of individual radionuclides released in gaseous effluents. Therefore, Detroit Edison must determine the quantity released of the individual radionuclides. 3.5.1 Quantifying Releases Using SPING Noble Gas Monitor The quantification of gaseous effluents (noble gases) is based on the Continuous Radioactivity Monitor on each of the Ventilation System release points. This monitor measures the gross radioactive material concentration in the effluent but not the individual radionuclides concentrations. The latter are required for evaluation of release rate, alarm setpoints, and cumulative doses. As required by TS 3.11.2.1, a gas sample is collected at least monthly from each of the six gaseous release points (Reactor Building Exhaust Plenum, Standby Gas Treatment System, Radweste Building, Turbine Building, Onsite Storage Facility, and Service Building). As discussed in ODCM Section 3.2.2, this gas sample is analyzed by gamma spectroscopy to identify principal gamma-omitting radionuclides (noble gases). The results of the sample analysis are used to determine the radionuclides distribution (i.e., fraction of total activity for each measured noble gas). For Containment PURGENENT, samples are collected prior to the initiation of the release and periodically throughout the release (see ODCM Section 3.2.1). These samples and analysis are used for correlating the radionuclides distribution with the SPING Monitor readings for determining total release. For en extended PURGENENT period (e.g., longer than 48 hours), drywell airborne activity levels will equilibrate. After equilibrium is reached, the quantification of the PURGENENT can be adequately addressed by the periodic (typically weekly) sample and analysis of the Reactor Building Exhaust Plenum or Standby Gas Treatment System. e
' ODCM-3.0 R:visi:n 2 P;ge 3.0-8 Based on the average Noble Gas Monitor reading over the' release period, the individual noble gas radionuclides releases are quant'fied by the equation: Og = 1.0 E + 03
Al

C

VF

T (3-8)
Where: Oi = total activity released of radionuclida I (uCl) Ai = activity of radionuclides i from the gamma spectral analysis of the grab sample from the release point (uCl) i C = average gross activity concentration over the release period as measured by the SPING Noble Gas l Monitor (uCl/cc) - VF = Ventilation System flow rate (liters / min) T = total time of the release period (min) f 1.0 E + 03 = milliliters per liter i The Eberline CT-28 Control Terminal for the SPING Monitors records i historical averages (e.g., daily averages for current 24 day history) of the release rates for all the SPING Monitors. By entering the time period of interest (e.g., release period), the Average Concentration C used in Equation (3-8) can be obtained from the CT-28 Terminal. The Release Period T is the time from the last sample of the release point to the time of the current sample under evaluation. j As required by TS Table 4.11.2.1.2-1, special samples are required of the a RB Exhaust Plenum and SGTS following shutdown, startup or a THERMAL POWER change exceeding 15% within a 1 hour period. Exceptions to this special sampling are allowed as noted previously in ODCM Section 3.2.2. Equation (3-8) can be used for quantifying the releases based on this special sample analysis. The specified release period in this situation is shorter than the typical 7 days. However, the release period still is represented by the time from the last sample to the time of i the current sample under consideration. if no activity is detected in the gas sample (i.e., all activity less than Lower Limits of Detection (LLD)). the default radionuclides distribution of Table 3-1 should be used unless better data is available to substantiate a different I distribution. Until a radionuclides distribution can be established for Fermi 2 based on actual operating conditions and measured effluent, the design distribution of UFSAR Section 11.3 should be used. This distribution is presented in Table 3.0-1.
,g Page 3.0-9 3.5.2, Ou:ntifying Role:sa Rate and Total R;I;;s:s with Monitor inop r213 The requirement to quantify radioactive releases continues even while monitors are incperable. In this situation, a backup method is used to relate effluent concentrations and volumes to a total radioactive material release. Analysis of grab samples provides the radioactive material concentrations in the effluent. The flow measurement device, or flow estimate, and the release duration provide the total volume released. With these, the backup method can determins the release rate and resultant total ainount of radioactive material releassd. 1. Release Rate Evaluation With an inoperable monitor, the demonstration of compliance with.the release rate limit of TS 3.11.2.1a must be based on the periodic grab samples. These grab semples provide a measurement of the noble gas concentration in the effluent stream. Equations (3-4) and (3-5) can be used for determining the dose rate based on the grab samples. For the applicable monitor and release point, the calculated dose rates should be compared with the allocated dose rates of Table 3.0-2. If the calculated exceeds the allocated, an evaluation of the total dose rates for the site should be performed. The dose rate for each release point should be determined and these contributions summed to determine the total dose rate for the site. I 2. Total Release Evaluation When a SPING Noble Gas Monitor is inoperable, operators lose the capability to determine the average release rate over the release period of interest. The periodic grab samples must be used to quantify the total releases. The measured noble gas radionuclides concentrations in the grab samples will be considered representative of the average effluent concentration over the period since the last sample. The following equation may be used for determining the release quantities j from any release point based on the grab sample analysis: Og = 1.0 E + 03
WT

T

C1 (3-9)
Where: Oj = total activity released of radionuclides I (uCI) . VF = Ventilation System release rate (liters / min) T = total time of release period (min) 1.0 E + 03 = mifilliters per liter C; = concentration of radionuclides i as determined by gamma spectral analysis of grab sample (uCi/cc) I I j

~~_----___J~~

WevtsUon N Page 3.0-10 i 3.6 Site Csundary Do30 Rata - Radiolodins cnd Particulat:s TS 3'.J1.2.1.b limits the dose rate to _$1500 mrom/yr to any organ for I-131,1-133, tritium and particulate with half-lives greater than 8 days. To demonstrate compliance with this limit, an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period (nominally once per 7 days). The following equation may be used for the dose rate evaluation: (R
06) bo = X/O

i (3-10)
Where: J Do = average organ dose rate over the sampling time period (mrom/yr) = atmospheric dispersion to the corg) rolling SITE BOUNDARY lo X/O for the inhalation pathway (sec/m from Table 3-4 3 = dose parameter for radionuclides I,(mrom/yr per uCi/m ) for the RJ child inhalation pathway from Table 3-5 U = average release rate over the appropriate sampling period and i analysis frequency for radionuclides I -- I-131,1-133, tritium or other radionuclides in particulate form with half-life greater than 8 days (uCl/sec) Og = Ci
VF

1.67E + 01 Where:
VF = Average ventilation flow for release point (liters / min) = Concentration of radionuclides i as determined by gamma spectral C1 analysis of media (uCi/ml) 1E + 03 (cc/ liter) * (1 min / 60 sec) 1.67E + 01 = l 3.6.1 Simplified Dose Rate Evaluation for Radiolodines and Particulate It is conservative to perform a simplified evaluation of allowable releases by applying the I-131 dose factor to the collective releases for all measured radionuclides. By substituting 1500 mrom/yr for 6o and solving for 61, an allowable release rate can be determined. Based on the annual average meterological dispersion (see Tab le 3.0-4) and ' the dose factor for the most limiting potential pathway, age group and3 organ (inhalation, child, thyroid -- R = 1.62 E + 07 mrem /yr per uCl/m ), i l' the allowable release rate (based on 1-131)is 1g uCi/sec. For a 7-day j period, which is the nominal sampling and analysis frequency, the cumulative release would be 11 Cl. Therefore, as long as the collective releases in any 7-day period do not exceed 11 Cl, no additional analyses are needed to verify compliance with the TS 3.11.2.1.b limits on allowable rolesse rate. m.__
Page 3.0-11 3.7 Noble Gas Effluent Dose GCiculati2ns - 10 CFQ 50 3.7.1 - UNRESTRICTED AREA Dose - Noble Gases TS 3.11.2.2 requires e periodic assessment of releases of noble gases to evaluate compliance with the quarterly dose limits of 5 mrad, gamma-air and 10 mrad, beta-air and the calendar year limits 10 mrad, gamma-air and 20 mrad, beta-air. The following equations may be used to calculate the gamma-air and beta-air doses: (M
~~Q )~~
~~D = 3.17 E - OS~~
X/O

i i
~~7 (3-11) and (N~~
0)
~~= 3.17 E - OS~~
X/O

i 1
(3-12) Where: = sir dose due to gamma emissions for noble gas radionuclides (mrad) = air dose due to beta emissions for noble gas radionuclides (mrad) = atmospheric dispersion to the cy) trolling X/O SITE BOUNDARY location (sec/m Og = cumulative release of noble gas radionuclides i over the period of interest (uCl) = sir dose factor due to gamma ogssions from noble gas Mi radionuclides I (mrad /yr per uCi/m, from Table 3.0-3) l Ni = sir dose factor due to beta emissgons from noble gas radionuclides i (mrad /yr per uCi/m. Table 3.0-3) 3.17 E - 08 = 1/3.15 E + 07 (year /sec) 3.7.2 Simplified Dose Calculation for Noble Gases j in lieu of the individual noble gas radionuclides dose assessment presented above, the following simplified dose calculational equations may be used for verifying compliance with the dose limits of Technical Specification 3.11.2.2. (Refer to Appendix C for the derivation and justification of this simplified method.) D = 2.0
3.17 E - 08

X/O

Moff

Oi 7
~~(3-13) and = 2.0~~
3.17 E - OS

X/O

Neff

Oi (3-14)
ODCM-3.0 Revislan 2 Page 3.0-12 Where: = 2.7 E + 03, effectivg) gamma-air dose factor Met -(mrad /yr por uCi/m { l = 2.3 E + 03, effectivg) beta-air dose factor Ngf (mrad /yr per uCl/m 2.0 = conservatism factor to account for potential verlebility in the radionuclides distribution 3.8 Radiolodine end Particulate Dese Calculations - 10 CPR 50 3.8.1 UNRESTRICTED AREA Dese - Radiolodine and Particulate in accordance with requirements of TS 3.11.2.3, a periodic assessment is required to evaluate compliance with the quarterly dose limit of 7.5 mrom : and the calender year limit of 15 mrom to any organ. The following equation may be used to evaluate the maximum organ dose due to releases l of I-131, tritium and particulate with half-lives greater than 8 days: Deep = 3.17 E - 88
W

SFp*
~~(P4~~
~~Og)~~
(3-15) Where: Daop = dose or dose commitment via controlling Pathway p and Age Group a (as identified in Table 3.0-4) to Organ o, including the total body (mrom) W = atmospheric dispersion parameter to the controlling location (s) as identified in Table 3.0-4: I W =* X/Q atmospheric dispersion for inhalation pathwy) and H-3 dose contribution via other pathways (se::/m = D/Q, atmospheric deposition for vegep) tion, milk and W ground plane exposure pathways (m' Where: 3 l Rg = doge factor for radionuclides I, (mrom/yr per uCl/m ) or (m - mrom/yr por uCl/sec) from Table 3.0-5 for each Age Group (a) and the applicable Pathway (p) as identified in Table 3.0-4. Values for Rg were derived in accordance with the methods tiescribed in NUREG-0133. l Og
~~cumulative release over the period of interest for~~
' radionuclides I - I-131 or radioactive motorial in particulate form with half-life greater than 8 days (uCI). = ennual seasonel correction factor to account for the ( SFp fraction of the year that the applicable exposure pathway does not exist:
- QcvCs0cn 3 ~ Page 3.0-13 1
~~1) For milk and vegetation exposure pathways:~~
~~= A six monthefresh vegetation and grating season (May through October) limits exposure through i this pathway to half the year = 0.5 (derived from flog Guide 1.10g, Rev 1)~~
~~2) For Inhalation and ground plane exposure pathways:~~
= 1.0 (derived from Reg Guide 1.10g, Rev 1) 3.17 E - 08 = 1/3.15 E + 07 (year /sec) The age group with the highest potential dose via the controlling pathway-a should be used for evaluating the maximum exposed individual. This I determination is based on a comparison of the age group pathway dose conversion factors (Table 3-5). The infant age group is controlling for the milk pathway and the child age group is controlling for the vegetable pathway. Only the controlling age group and pathway identified in Table 3.0-4 need be evaluated for compliance with TS 3.11.2.3. 3.8.2 Simplified Dose Calculation for Radioiodines and ParticuGetes in lieu of the individual radionuclides (I-131 and particulate) dose assessment presented above, the following simplified dose calculation may be used for verifying compiduce with the dose limits of TS 3.11.2.3. Dmax
I17 I - 88

W

SFp

Re-1g;

Og (3-16) f Where:
Dmax = maximum organ dose (mrom) R -131 = l-131 dose parameter for the thyroid for the identified l controlling pathway = 4.70 E + 10, child thyrgld dose parameter for the vegetable pathway (m - mrom/yr per uCi/sec) The ground plane exposure and inhalation pathways need not be considered when the above simplified calculational method is used because of the oven.Il negligible contribution of these pathways to the total thyroid dose. It is recognized that for some particulate radionuclides (e.g., Co-60 and Cs-137), the ground exposure pathway may represent a higher dose j-contribution than olther the vegetation or milk pathway. However, use of the I-131 thyroid tiose parameter for all radionuclides will maximize the organ dose calculation, especially considering that no other radionuclides has a higher dose parameter for any organ via any pathway than 1-131 for the thyroid via the vegetable or milk pathway.
Rovisisn 2 Page 3.0-14 The location of exposure pathways (critical receptors) and the corresponding maximum organ dose calculation should be based on the pathways identified by the annual land-use census (Technical Specification 3.12.2). Otherwise, the dose should be evaluated based on the predetermined controlling pathways identified in Table 3.0-4. 3.9 Gaseous Effluent Dose Projection As with liquid effluents, the Fermi 2 TS on gaseous effluents require " processing" of gaseous effluents if the projected dose exceeds specified limits. This TS implements the requirements of 10 CFR 50.36a on maintaining and using the appropriate radweste processing equipment to keep releases ALARA. TS 3.11.2.5 requires that the VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when the projected dose exceeds 0.3 mrom to any organ in any 31 day period (i.e., one-quarter of the design objective rate). Figure 3-1 presents the gaseous affluent release points and the VENTILATION EXHAUST TREATMENT SYSTEMS applicable for reducing effluents prior to release. Dose projection is performed at least once per 31 days using the following equation: Dmamp = Dmax * (31/ d) (3-17) Where: Dmaxp = maximum organ dose projection for current 31 day projection (mrom) Dmax = maximum organ dose to date for current calendar quarter as determined by Equation (3-15) or (3-16) (mrom) d = number of days to date in current calender quarter 31 = number of days in projection END OF SECTION 3.0 l 1 l l l ]
P ge 3.0-15 TABLE 3.0-1 l Default Noble Gas Radionuclides Distribution
~~of Gaseous Effluents~~
' Radionuclides Fraction of Total (A / { Ag) i Kr-85m 0.10 Kr-85 0.01 Kr-88 0.04 Kr-89 0.06 Xe-133 0.67 Xe-135 0.02 i Xe-137 0.02 Xe-138 0.07 TO.'AL 0.99 NOTE: Data adapted from Form 12 UFSAR, Section 11.3, Table 11.3-5. Kr-g0, Kr-91, Xe-139, and Xe-140 have been excluded from the distribution. Because of their short half-lives, they decay during transport off site to negligible levels of activity. Kr-87, Xe-131m. and Xe-133m have been excluded because of their negligible fractional abundance.
ODCM-3.0 Revision 2 '900e 3.0-16 ./ TABLE 3.0-2 Generic Values for Evaluating Gaseous Release Rates and Alarm Seapoints Allocation Allocated Dose Release Point Flow Rate Factor Este Umit Generic Alarm (liter / min) (AF) (mrom/ year) Setpoint (uCi/ml) R: actor Bul!d!ng - 2.67E6 0.50 T Body = 250 1.02E-4+ Skg. Exhaust Plenum Skin = 1500 D11-P280 Organ = 375 Stendby Oss 1.07ES 0.10 T Sody = 25 8.12E-4+ Skg Trcatment System Skin = 150 Div i D11-P275 Organ = 75 Standby Gas 1.12E5 0.10 T Body = 25 6.17E-4+ Bkg Treatment System Skin = 150 ' Div ll D11-P276 Organ = 75 Turbine Building 8.67E6 0.20 T Body = 50 1.06E-5+ Bkg V:ntilation Skin = 300 D11-P279 Organ = 150 Service Building g.06E5 0.01 T Body = 2.5 7.g3E-6+ Bkg V:ntilation Skin = 15 D11-P282 Organ = 7.5 Radweste Building 1.13E6 0.02 T Body = 5 6.22E-6+ Bkg V:ntilation Skin = 30 D11-P281 Organ = 15 Onsite Storage 3.06ES 0.02 T Body = 5 1.g3E-4+ Bkg Skin = 30 Building V:ntilation Organ = 15 D11-P281 Q actor Building 2.41E5 0.50 T Body = 125 g.1gE-6+ Bkg Skin = 750 V:ntilation* Culf Atomic M:nitors 011-N408,N410 D11-N408 and N410 will start the SGTs, close the Drywell PurgeNent valves, isolate Rx Building Ventilation System, isolate Control Center, and initiate emergency recirculation mode. Alarm setpoints for these monitors are not required by Fermi 2 TS but have been included in this table for completeness.
Papa 3.0-17 TABLE 3.0-3 Dose Factors for Noble Gases
Total Body Skin Gamma Air Beta Air Gamma Dose Beta Dose Dose Factor Dose Factor Nuclide Factor Ki Factor Li Mi Ni (mrom/yg)per (mrom/yg)per (mrad /yrg)er (mrad /yrg)er uCl/m uCl/m uCl/m uCi/m 1.93E+01 2.F8E+02 Kr-83m 7.56E-02 Kr-85m 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 134E+03 1.72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 1.47E+04 237E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1.73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe-131m 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.4BE+03 Xe-133 2.94E+ 02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3.12E+03 7.11E+02 336E+03 739E+02 Xe-135 1.81E+ 03 1.86E+03 1.92E+03 2.46E+03 Xa-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E <03 4.13E+03 9.21E+03 4.75E+03 Ar-41 8.84E+ 03 2.69E+03 930E+03 3.28E+03 NOTE:
Dose factors taken from NRC Regulatory Guide 1.109
CDCM-3.0 Revision 2 Paga 3.0-18 7ABLE 3.0-4 Controlling Locations, Padtways, and Atmospheric Dispersion for Dose Calculations
~~Atmospheric Dispersion Factor Technical Location Pathway (s)~~
Controlling X/O D/O I 2 Sp;cification Age Group (sec/m ) (1/m ) 3.11.2.1s site boundary noble gsses N/A 4.94E-6 N/A (0.915 km,NW) direct exposure 1 3.11.2.1 b site boundary inhalation child 4.94E-6 N/A (0.915 km,NW) 3.11.2.2 site boundary gamma-sir N/A 4.94E-6 N/A (0.915 km,NW) beta-air 3.11.2.3 residence
~~milk, infant 2.26E-7 9.11E-10 (3.379 km,WNW) inhalation, and ground plane NOTE:~~
The identified controlling locations and pathways have been determined from the 1988 land-use census data. The atmospheric dispersion factors for these locations were derived from meteorological data records for the period 1/1/87 to 12/31/87.
s.wwe Er:1.eu oeian Doce Goemineoc Utmom -Page 3.0-19 s... se631:sses rzews, s.n resp)re = eent (tses/tr per ots/s 62.LL1 T.Ded?. Seelses teos &&wr T6trass Egeosy &ses - ~~ ~~ 3.36Ee3 1.36t.3 1.365 1 1.261 3 1.16t.3 1.368 3 33 C.le 1.82t+e 3.etE 3 3.elt+3 3.611 3
3. ele +3 3.,611+ 3

3. 44 t + 3 se.24 1.02E+6 1.02t+e 4.02E+e 1.021 6 1.03E+6 1.021 6 1.E2t+6 S'
~~0.64E+4 S.011+6 9 32 1.323 6 f.71t+4 3.93E+1 3.20E*1 1 6e E+e 3.323 3 1.00t+2 Ct.S! 9.66E+1 3.408 6 7.168 4 6.30t+3 3.96t+e me.La~~
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1. ht E+ 5 p.14 h.163 1.38t+? 3.63t+3 La.160 1.978 3 9.92t+2 e.645 1 1.b8E.S be.142 1.603 4 4.358 3 B. dst +8 1.81E+6 6.173 +4
.let 1.96t+5 1 13E.5 3.778e? 8.31t+1 3.368+3 463 1.00$+3 7.48tel 1.11t+10 1.775 6 8.165+6 Co.164 3.395e? 1 388+? 3.?85+8 3.168 3 1.47864 Pr.143 6.34 E+4 3.54 E+4 Pr.16e 3.35E+4 3.85t+8 3.333 3 56 167 3.54E+4 3.86E 4 1.0SE*? 1.11E 6 u.187 3.838+e 3.89t+4 3.815+? ?.?IE+1' l 89 239 1.638 3 1.608 3 4.375+2 + 4 1
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1. De t + ? 4.30t+4 Cr.Sl 9.37t+8 8.911+7 1.3St+8 4.88t+8 No.Se

9. Set +2 3.981 0 1.638+1 1.655 1 No.54 1.668+8 9.98te? 8.38t+1 fe.SS 3.388 8 3.385+8 1.33E+8 9.98t+8 1.631 8 Fe.89 1.01B+8 4.33t+8 3.968+8 S.005+7 1.?98+?
Co.S? 6.968 0 1.91t+8 4.388+7 Co.58 3.36t+9 S.60t+8 3.494+8 Co.60 8.81t+8 S.eSt+8 31 43 1.61t+10 1.13t.9 3.978 3 3.33t+0 l es.65 S.73t+1 1.335 0 6.815 S 3.95t+3 3.13t+4 8.685 3 Co.44 6.135 8 6.D6t+8 9.41B+8 Bo.45 4.848 0 1.478 9 3.888 5 1.091 6 t 1.038 5 b 49 S.198 6 1.964 3 = 3.338 6 k 83 3.088 0 - 83 44 Dr.85 e.988+? 1.383*8 3.738+4 86 66 36 48 86 89 1.08t+9 4.38t+8 Br.89 1.51t+10 3.11t+10 1.0St+11 Br.90
~~7. Sit +11 1.36t+6 1.19t+e Sr.91 3.99t.S 1.0!!+4 1.69t+1 Br 93 3.97t+3 1.03E+8 3.36t+3 7 90 1.36t+4 3.961 7 7 91e 9.438 9 3.3St+9 3.11E+$~~
~~7 91 f.8?t+6 3.335 6 3.45E.3 ?.93 8.678 1 e.988 6 a.67t+0 7 93 1.638 3 1.3?t+9 3.785.S 8.078+5 tret$~~
~~1. tee +6 S.698+5 1.688+1 3.81t+1 9.36t+1 3r.9?~~
~~3.09t+3 6.11t+1 4.35t+8 S.D6t+4 1.03E+5 Ob.95 1.93t+5 1.06t+$ 1.998 3 3.648.?~~
~~1. DOE.?~~
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~~4. bee +?~~
~~Te.13?e S.Sitet 1.96 t+8 1.318 8 3.34B+9 4.19t+5 1.178+3 To.13? 8.438 3 1.93t+3 3.74E+3 3.308 4 1.30t+9 S.88t+1 7e.13 De 3.6?t+8 1.34E+8 1.18t+8~~
~~1. Set +9 3.605 3 1.318 To.839 6.138 4 3.33 t.e 4.458 4 3.615 3 3.138 7 3.34tel To.138e 0.648+S 6.998+S 6.998+5 4.338 6 To.131 7.838.? 3.33t+6 To.133 3.90t+6 3.478+6 3.60t+6 3.37t+?~~
1.07t+$ e.09t+S 1 130 3.96t+S 3.038+6 8.35t+7 1.98t+4 3.13E+? 3.?tt+? I.131 f.70E+1 1.088 8 3.14E+10 1.05E.8 S.91t+1 4.07t+1 8 183 S.188+1 8.968 3 4.878+3 3.168+3 3.838 4 1.93E+6 3 133 1.978 6 3.34E+4 a.46t+6 S.068 4 3.3St.4 9.138 5 9.998 5 3.348 6 4.348 3 4.018 4 1.lle 1.05t+S 3.$3t+a 135
3. 688+e

9. 68 t +4 6.198 6 1.50t+5 S.30t+9 3.03t+9 3.088*8 f.74E+9 Co.134 f.998+9 1.67t+10 9.198+7 1.458+? 1.36t+1 1.13E+8 Co.136 6.398+7 1.098+8
= e.D98+9 1.?st+9 1.988 4 4.698 9 Co.137 1 015+10 1.388 10 go.138 8.048.S 1.368 8 3.478 1 S.088 4 to.139 8.??S.3 1.988 9 S.?Stee 1.168 5 3.138 4 S.918+6 to.160 1.388*8 1.80E+S be.le! to.143 S.08t+1 3.3St+3 be.160 1.00E+3 8.968+2 1.78t*0 1.638 5 be.843 1.388 4 S.698 5 S.388+8 3.16t+4 0.065 4 4 161 ).838+l 1.08tes 3.95t+1 7.43t+1 3.068 3 b los v.3?t+3 6.83 8+5 1.33B+10 3.838 4 1.388+? Co. lee S.378+7 8.188+? 3.368 4 3.S$3+3 1.688+4 Pr.143 ?.138+4 3.068 4 = pr. lee 1.48t+8
~~d. bet +3 3.33E+4 58 467 S.638*4 3.968 4 7.86t+6 1.03E+4 v.18?~~
~~3.358 4 3.90s.6 3.105 7 f.34 ten 4.998+3 i Op.339~~
~~1. 988+ 3 1.308+3 I~~
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1.De87 See1883. Doe) Liege T6frett 818ee ?. 8 4.Olt+3 6.DI E+ 3 4.SiE+3 4.91883 4.Olt+3 'e.08t+3 5 3-C le 3.h0E*6 7.01t+5 1.01E*S 1.Olt+5 7.01t+5 f.elt+5 ?.0!!eS se.34 3.03t+5 3.03tel S.83E+$ 3.83t+$ 3.83t+5 3.aSt+5 3.83t+5 ,e I 9.30t+7 2.30t+8 P.32 3.37t+9 1.98E 8 6.D6t+4 1.?9t+4. 1.191+5 S.39t+6 1.181 5 Ct.Sl i B.SSE*8 1.76t+8 1.85t+8 6.61E*8 me.Se 3.?St+3
~~4. 38t+0.~~
~~3.361+1 1.90E+ 1 me.56~~
~~3. hot *4 f.Det.? 3.31t+8 -~~
~~Fe.SS 8.00t*4 4.36t+8 1.085 8 6.76t+8 3.23t+8 Fe.59 4.015+4 6.69E*8 3.eSt+8 6.64E+1 3.995+? Co.S? 3.??t*8 1.988*8 6.67t+? Go.58 3.188 9 1.13t+9 3.98 5+8 Co.60 1.ett+8~~
~~1. 36 t + 9.~~
31 63 3.95t+10 3.118 9 1.315+3 S.??t+0 96 65 1.DSE+2 9.99 t +0 9.30s.3 6.69t+3 2.68t+4 1.llE*4 Co.64 3.00t+8 1.35t+9 1.368 9 Se.45 S.13E+4 3.16Ee9 3.308 3 3.93b.6 1.835 5 Se.69 1.318 5 3.185 5 3.DeE+6 Dr.82 3.88E,0 3r.83 Dr.44 Dr.85 8.91E+1 3.?SE+8 e.83E+4 86 86 86 88 86 99 1.398+9 1.038 9 St.89 3.998+10 1.67t+10 3.19t+11 l St.90 1.36t+12 3.38t+6 3.D81 4 Sr.91 S.50t+5 1.58t+e 3.92t+1 Sr.92 7.28t+2 6.56te? 6.175+2 - 7 90 3.30E+e 1.958 3 1 9te 9.968 9 3.69t+9 S.08t+5 7 91 1.8?t+7
~~6. 31 E+4 4.665 2 7 92 1.66t+0 4.68t+6 p.3St+0 i~~
~~J 7 93 3.Clt+2 3.951 8 7.641 5 3.33t+6 Ir.95 3.901 6 8.68t+5 1.13t+7 4.81L+1 1.175+2 3r.97 S.64E.2 0.lSt+1 3.95t+8 3.let+S~~
~~1. bot +5 86 9S 4.10t+5 1.D9t+5 3.738 3 4.13t=?~~
9.821 7 56 91 4.90E.6 8.851 7 6.488 6 1.968 6 3.4?t+? 7.83t+6 Re 99 1.335 2 4.638 0 S.19t+3 1.81E+2 1s 99e 4.488 0 9.12t+0 1s.103 3 998+8 S.Det+6 3.99te? to.103 1.S$te? S.90E.6 3.33t 1 8.068+2 Sm 105 9.l?t+I 1.168 10 9.30E+f 1.01E*9 to 106 7.65t+8 th-103e 36 106 3.b88+9 1.tetet 4.pSE*? 48.lles 3.328 7 3.!?t+? 1.96t+8 3.30t+9 1.33t+8 l $6 134 3.52t+8 4.S?t+6 1.78;+5 3.78t+4 3.39t+9 3.051 8 i $6 125 6.99t+8 3.tSt+6 4.628 5 3.38t+8 4.6?t+1 l T o.12 Se 3.Slt+8 9.60t+7 9.SeE+7 1.075+9 1.S?t+8 l 1e.117e 1.328 9
3. De B +8 3.16E+8 3.??l+9 3.91E+$ 3.39t+3 fe.137 1.001 4 3.70E+3 6.93t+3 3.05t.4 1.DeE+9 1.33t+8 To.139e 8.56E*8 3.998+8 3.75E+8 3.SIE+9 1.173 1. S.?et e To.139 1.158 3 3.338 4 8.338 6 3.373 3 3.168+? 8.68E+5 1e.13te

1. Del +6 S.33B+S 1.80E*4 S.865+4 fe 131 3.11t+7 3.735 6 fe.132 6.988 4.3.998+4 4.90E +6 3.878+7 2 130 8.318+5 1.36t*6 1.38t*4 1.88t+6 3.37t+$ 6.4?t+5 1.388 7 8.Itt+?
~~1 131 1.63B+4 1.64t+8 4.?68+10 3.348 8 1.99t+2 7.??!*1 9.30E+1 1.695+3 ?.848+3~~
~~3. Set +2~~
~~= 1132 1.195+6 1.688+6 133 3.995+4 4.e4E+6 S.388+8 9.40E+6 f.134 1.998 4 3.165 4 7.308 3 e.048 4 3.105 4 1.488 4 S.988 4 S.975+e E.135~~
6. 64 B +4 1.18t+l 1.04547

1. Sit +5 8.165+9 3.93E+9 1.635 8 S.Det+9 Co.lSe 1.688+30 3.685*10 1.198+8 1.965+t 7.998+6 1.43t+8 Co.136 S.088+7 3.83E+8 7.668+9 3.8st+9 1.438+8 3.388+9 Se 137 3.398+10 8.998+10 go.134 3 388.$ 1.618 5 3.95t+0 1.448 3 so.139

9. !!E.3 3.738 5 T.90E+4 1 6SE+S 1.40t+8 1.638+7 he.160 3.??t+4 3.638+5 to. net De 843 3.15t+7 3.91B+3 be.le0 3.23t+3 1.13t+3 1.ett+1 3.325 5 be.le2 3.338 6 f.608 3 1.64t+7 9.13t+3 3.698+4 341 1.pE+l $.54E+4 1.37t+7 1.96t+2 5.938 3
.l63 1.73t+3 v.36E+5 1.DeE+10 6.78t+4 3.31E+7 Ce=16e 3.375+8 3.988+? 1.40t+8 f.378+3 3.615+4 pr.le3 1.e4848 4.465+4 Pr. lee 9.18E+7 4.69t+3 3.18E+4 58 847 1.165+4 S.00 tee B.SSE*6 1.72E++ v.187 6.67E+4 3.838*4 1.96E*? 1.398 2. S.305 2 sp.339 3.888 3 1.83E I l
'Page 3.0-33 T;613 S.5 (s%tteeld) 843. Gr:ged F111e Pet 6ee? Does FIsttra (;
~~eres/yr, per oCs/Es)~~
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Co.364 6.95t+7 Pr 145 Pr.164 1.83E.3 54 167 8.60t+6 v.187 3.S6E.4 SP.239 1.78t+6 4
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R; vision 2 Page 4.0-2 ] As appropriate for demonstrating / evaluating compliance with the limits of Tcchnical Specification 3.11.4 (40 CFR Igo), the results of the environmental monitoring program may be used to provide data on actual measured levels of radioactive material in the actual pathways of exposure. ) 4.2.1 Effluent Dese Calculations For purposes of implementing the surveillance requirements of TS 3.11.4 and the reporting requirements of 6.g.1.8. dose calculations for Fermi 2 may be performed using the calculational methods contained within thir, ODCM; the conservative controlling pathways and locations of Table 3.0-4 or thw actual pathways and locations as identified by the land use census (TS 3.12.2 and ODCM Section 5.0) may be used. Uguld pathway doses may be calculated using Equation (2-10). Doses due to releases of radiolodines, tritium snd particulate are calculated based on Equation (3-15). The following equations may be used for calculating the doses to MEMBERS OF THE PUBLIC from releases of noble gases: Deb = 3.17 E - SS
X/O *
~~(K;~~
~~Og)~~
~~(4-1) and Ds = 3.17 E - OS~~
X/O *
~~([Li + 1.1 Md~~
~~Q )i (4-2) where:~~
Db = total body dose due to gamma emissions for noble gas t radionuclides (mrom) Ds = skin dose due to gamma and beta emissions for noble gas radionuclides (mrad) 3 X/O = atmospheric dispersion to the offsite location (sec/m ) 0; = cumulative release of noble gas radionuclides i over the period of interest (uCl) ~ = C x VF x 1.67E + 01 CJ = concentration of radionuclides i as determined by gamma spectral analysis of media (uCi/ml) VF = average ventilation flow for release point (liters / min) 1.67E + 01 = (IE + 03 ml/ liter) * (1 min /60 see) kg = total body dose factor due to gamma emissig)ns from noble gas radionuclides I (mrom/yr por uCl/m (from Table 3.0-3) Li = skin dose factor due to beta emisglons from noble gas I radionuclides I (mrom/yr per uCi/m ) (from Table 3.0-3) l l l l
ODCM-4.0 flevisisn 2 Page 4.0-3 = gamma air dose fag)or for noble gas radionuclides I Mi (mrad /yr per uCl/m (from Table 3.0-3) 1 1.1 = mrom skin dose per mrad gamma sir dose (mrom/ mrad) 3.17 E - 08 = 1/3.15 E + 07 yr/sec Average annual meterological dispersion parameters or meterological conditions concurrent with the release period under evaluation may be used (e.g., quarterly averages or year-specific annual averages). 4.2.2 Direct Exposure Dose Determination From evaluations performed in the Fermi 2 Environmental floport, Section 5.3.4, the direct exposure to the highest offsite location from the Turbine Building N-16 skyshine dose has bosn calculated to be approximately 3 mrom/ year. This value may be used as a baseline for actual direct exposure contributions during plant operations. Other potentially significant direct esposure contributions to offsite individual doses may be evaluated based on the results of the environmental measurements (e.g., TLD, ion chamber measurements) or by the use of a radiation transport and shielding calculational method. Only during atypical conditions will there exist any potential for significant onsite sources at Fermi 2 that would yield potentially significant offsite doses to a MEMBER OF THE PUBUC. However, should a situation exist whereby the direct exposure contribution is potentially significant, onsite measurements, offsite measurements and calculational techniques will be used for determination of dose for assessing 40 CFR 100 compliance. The calculational techniques will be identified, reviewed, and approved at that time. 4.2.3 Dose Assessment Based on Radiological Environmental Monitoring Data Normally, the assessment of potential doses to MEMBERS OF THE PUBLIC must be calculated based on the measured radioactive effluents at the plant. The resultant levels of radioactive material in the offsite environment are so minute as to be undetectable. The calculational methods as presented in this ODCM are used for modeling the transport in the environment and the resultant exposure to offsite individuals. The results of the radiological environmental monitoring program can provide input into the overall assessment of impact of plant operations and radioactive effluents. With measured levels of plant related radioactive material in principal pathways of exposure, a quantitative assessment of I potential exposures can be performed. With the monitoring program not identifying any measurable levels, the data provides a qualitative assessment - a confirmatory demonstration of the negligible impact. 6ose modeling can be simplified into three basic parameters that can be applied in using environmental monitoring data for dose assessment: D = C
U

DF (4-3) l t
Revisign 8 P:g) 4.0-4 wh;re: D = dose or dose commitment C = concentration in the exposure media, such as air concentration for the inhalation pathway, or fish, vegetation or milk concentration for the ingestion pathway U = individual exposure to the pathway, such as br/yr for direct exposure, kg/yr for ingestion pathway DF = dose conversion factor to convert from an exposure or uptake to an individua! dose or dose commitment 3 The applicability of each of tr.ese basic modeling parameters to the use of environmental monitoring data for dose assessment is addressed below: Concentration - C The main value of using environmental sampling data to assess potential doses to individuals is that the data represents actual measured levels of radioactive material in the exposure pathways. This eliminates one main uncertainty and the modeling has been removed - the release from the plant and the transport to the environmental exposure medium. Environmental samples are collected on a routine frequency (e.g., weekly airborno particulate samples, monthly vegetable samples, annual fish ~ samples). To determine the annual average concentration in the environmental medium for use in assessing cumulative dose for the year, an average concentration should be determined based on the sampling frequency and measured levels: $= {(Cg
~~t)/365 (4-4) where:~~
k = average concentration in the sampling medium for the year = concentration of each radionuclides i measured in the C3 Individual sampling medium t = period of time that the measured concentration is considered representative of the sampling medium (typically equal to the sempling frequency; e.g.,7 (lays for weekly samples,30 days for monthly samples). If the concentration in the sampling medium is below the detection capabilities (i.e., less than Lower Umits of Detection (LLD). a value of zero should be used for C (Ci = 0). 1
ODCM-4.0 Qevision 2 ~ Page 4.0-5 Exposure - U Default Exposure Values (U) as recommended in Regulatory Guide 1.109 are presented in Table 4.0-2. These values should be used only when specific date applicable to the environmental pathway being evaluated is unavailable. s Also, the routine radiological environmental monitoring program is designed to sample / monitor the environmental media that would provide early indications of any measurable levels in the environment but not necessarily levels to which any individual is exposed. For example, sediment samples are collected in the area of the liquid discharge: typically, no individuals are directly exposed. To apply the measured levels of radioactivity in samples that are not directly applicable to exposure to rest individuals, the opproach recommended is to correlate the location and measured levels to - octuallocations of exposure. Hydrological or atmospheric dilution factors can be used to provide reasonable correlations of concentrations (and doses) at other locations. I The other alternative is to conservatively assume a hypothetical individual at the sampling location.- Doses that are calculated in this manner should be presented as hypothetical and very conservatively determined - actual exposure would he much less. Samples collected from the Monroe water supply intake should be used for estimating the potential drinking water doses. Oth2r water samples collected, such as near field dilution area, are not applicable to this pathway. ' Dose Factors - DF The dose factors are used to convert the intake of the radioactive material to en individual dose commitment. Values of the dose factors are presented in NRC Regulatory Guide 1.109. The use of the RG 1.109 values applicable to the exposure pathway and maximum exposed individual is referenced in Table 4.0-2. Assessment of Direct Exposure Doses " Thermoluminescent Dosimeters (TLD) are routinely used to assess the direct exposure component of radiation doses in the environment. However, because routine releases of radioactive material (noble gases) are so low, f the resultant direct exposure doses are also very low. A study
performed for the NRC concluded that it was generally impractical to distinguish any plant contribution to the natural back0round radiation levels (direct exposure) below around 10 mrom per year. Therefore, for routine releases from nuclear power plants the use of TLD is mainly confirmatory - ensuring actual exposures are within the oppocted natural background variation.

NUREG/CR-0711, Evaluation of Methods for the Determination of X-and Gamma-Ray Exposure Attributable to a Nuclear Facility Using Environmental TLD Measurements, Gall dePlanque, June 1)79, USNRC.
ODCM-4.0 Revision 2 P;ge 4.0-6 ,[' For releases of noble gases, environmental modeling using plant measured releases and atmospheric transport models as presented in ODCM Section 3.6 and 4.2.1 represents the best method of assessing potential environmental doses. However, any observed variations in TLD measurements outside the norm should be evaluated. END OF SECTION 4.0 9 ______________________________w
ODCM-4.0 C:visisn 2 P ge 4.0-7 TABLE 4.0-1 $...... Assumptions for Assessing Doses Due to.. Activities inside SITE BOUNDARY
1.
$ 9 t 9 9.9 tco Fishing Visitor's Center Distance / Direction: 470 meters / E 470 meters / SSW Estimated Exposure 240 hr/yr 4 hr/yr Time: (20 hr/ week over (4 hr/ visit,1 visit 3 month period) per year) Exposure Pathways: direct exposure direct exposure (noble gases) (noble gases) Inhalation inhalation (H-3.1-131, -133, (H-3,1-131, -133 particulate) particulate) Meteorological ennual average annual average Dispersion: (as determined for (as determined for year being evaluated) year being evaluated) 3 3 1.93E-5 sec/m 5.74E-6 sec/m
~~Meterological data is providad from the monitoring year 1987.~~
~~ODCM-4.0 Revisi:n 2 P;ge 4.0-8 TABLE 4.0-2 Recommended Exposure Rates in Lieu of Gite Specific Data~~
Table Reference Exposure Pathway Maximum Exposed Exposure Rates for Dose Factor Age Group from RG 1.109 Liquid Releases Fish Adult 21 kgly E-11 Drinking Water Adult 7301/y E-11 Bottom Sediment Teen 67 h/y E-6 Atmospheric Releases 3
inhalation Teen 8,000 m /y E-8 Direct Exposure .All 6,100 h/y'* N/A Leafy Vegetables Child 26 kg/y E-13 Fruits, Vegetables Teen 630 kg/y E-12 and Grain Milk infant 3301/y E-14 1 Adapted from Regulatory Guide 1.109. Table E-5 Net exposure of 6,100 h/y is based on the total 8760 hours per year adjusted by a 0.7 shielding factor as recommended in Regulatory Guide 1.109. END O
O Nuclear Production - Fermi 2 ODCM-S.0 Offsita De:o Calcula: Sin Manu:1 Routslin 2 Pese s.0-1 ASSESSMENT OF LAND USE CENSUS DATA S.0 ASSESSMENT OF LAND USE CfdNSUS DATA A Land Use Census (LUC) is conducted annually in the vicinity of the Fermi 2 she. This census futfills two main purposes: 1) Meet requirements of TS 3.12.2 for identWylng controlling location /pethway for dose assessment of TS 3.11.2.3; and 2) provide date on actual exposure pathways for essessing realistic doses to MEMBERS OF THE PUBLIC. 5.1 Land Use Census os Rettuired by TS 3.12.2 I As required by TS 3.12.2, a tend use census shall be conducted during the growing season et least once per twelve months. The purpose of the census is to identify within a 5 mile distance the location in each of the 16 meterological sectors of the nearegt milk animal, nearest residence and nearest garden larger than 500 ft produ:Ing broadleaf vegetatinn. The data from the LUC is used for updating the location / pathway for dose assessment and for updating the Radiological Environmental Monitoring Program. If the census identifies a location /pethway(s) yielding a higher potential dose to a MEMBER OF THE PUBLIC than currently being assessed as required by TS 3.1* 2.3 (and ODCM Section 3.7 and Table 3.0-4), this new location pathway (s) shall be used for dose assessment. Table 3.0-4 shall be updated to include the currently identified controlling location / pathway (s). Also, if the census identifies a location (s) that yields a calculated potential dose (vis the same exposure pathway) 20% greater than a location currently included in the Radiological Environmental Monitoring Program, the new location (-s) shall be added to the program within 30 days. The sampling location (s), excluding control locations, having the lowest calcultted dose may be f deleted from the program after October 31 following the current census. As required l by TS 3.12.2 and 6.g.1.8, the new location /pethway(s) shall be identified in the next Semiannual Radiation Effluent Release Report. The following guideline shall be used for assessing the results from the land use census to ensure compliance with TS 3.12.2. 5.1.1 Data Complistion 1. Compile allocations and pathways of exposure as identified by the land use census. 2. From this complied data, Identify any changes from the previous year's consus, identify the current controlling location / pathway used in the ODCM Table 3.0-4. Also, identify any location / pathway currently included in the itEMP (Table 6-1). ARMS - N6 FORMATION SERVICES Date approved: Release authorized by: ChanDe numbers incorporated:,, LCR 88-032-ODM DSN Rev 2 Date DTC TMPLAN File 1715.02 Recipient
CDCM-5.0 Revisinn 2 P ge 5.0-2 3. Determine the historical, annual everage meterological dispersion parameters (X/O, D/Q) for any new location (i.e., location not previously identified and/or evaluated). All locations should be evaluated against the same historical meterological data set. 5.1.2 Relative Dese Significance For all now locations, calculate the relative dose significance by applicable pathways of exposure. 1. Relative dose calculations should be based on a generic radionuclides ' distribution (e.g., Fermi 2 UFSAR gaseous effluent source term or past year actual effluents). An 1-131 source term dose may be used for assessment of the maximum organ ingestion pathway dose because of its overwhelming contribution to the total dose relative to the other particulate. 2. The pathway dose equations of the ODCM should be used. 5.1.3 Data Evaluation 1. By exposure pathway, formulate a listing of locations. The listing should be in descending order of relative dose significance. Include the relative dose significance in the listing. 2. Verify the controlling location used in the ODCM Table 3.0-4. If any location / pathway (s) is identified with a higher relative dose, this location / pathway (s) should replace the previously identified controlling location / pathway in Table 3.0-4. If the previously identified controlling pathway is no longer present. the current controlling location / pathway i should be determined. 3. This listing should be used in the evaluation for revisions to the REMP and Section 6.0 of the ODCM in accordance with TS 3.12.2, Action item b. 4. Any changes in either the controlling location / pathway (s) of the ODCM dose calculations (Section 3.7 and Table 3.0-4) or the REMP (ODCM i Section 6.0 and Table 6-1) shall be reported to NRC in accordance with TS 3.12.2 Action items a. and t:. and TS 6.g.1.8. NOTE: As permitted by footnote to TS 3.12.2, broadleaf vegetation sampling may be performed at the SITE BOUNDARY in two' locations,in different sectors with highest predicted D/Qs, in lieu of the garden census. Also, for cont.orvatism in dose assessment for compliance with TS 3.11.2.3 (ODCM Section 3.7 and Table 3.0-4), hypothetical exposure location / pathway (s) may be assumed (e.g., milk cow at 5 mile location or garden at SITE BOUNDARY in highest D/O sector). By this approach, the ODCM is not subject to frequent revision as pathways and locations change from year to year. A verification that the hypothetical pathway remains conservative and valid is still required. Also, for NRC
~~ ODCM-5.0 Revision 2 P g) 5.0-3 reporting, the actual pathways and doses should be reported along with the hypothetical. The reporting of the actual pathway and doses provides a formal documentation of the more realistic dose impact. 5.2 Land Use Consus to Support Realistic Dose Assessment The LUC provides data needed to support the special dose analyses of the ODCM Section 4.0. Activities inside the SITE BOUNDARY should be periodically reviewed for dose assessment as required by TS 6.9.1.8 (ODCM Section 4.1). Assessment of realistic doses to MEMBERS OF THE PUBLIC is required by TS 3.11.4 for demonstrating compliance with the EPA Environmental Dose Standard,40 CFR 190 (ODCM Section 4.2). To support these dose assessments, the LUC shall include (a) areas within the SITE BOUNDARY that are accessible to the public; and (b) use of Lake Erie water on and near the site. The scope of the LUC shall include the following: 5.2.1 Assessment of areas onsite that are accessible to MEMBERS OF THE PUBLIC. Particular attention should be given to assessing exposure times for ice fishing near the Fermi 2 shoreline and visits to the Fermi 2 Visitor's Center. Data should be ut.ed for updating ODCM Table 4.0-1. 5.2.2 Data on Lake Erie use should be obtained from local and state officials. Reasonable efforts shall be made to identify individual irrigation and potable water users, and industrial and commercial water users whose source is Lake Erie. This data is used to verify the pathways of exposure used in ODCM Section 2.5. END OF SECTION 5.0 .i _m
ODCM-6.0 Nucl::r Production - Fcrmi 2 flowlslan 2 Offsito Do30 Calculati:n Manur.1 PeBe 6.0-1 flADIOLOOiCAL ENVlh0NMENTAL MONITORING PROGRAM e 6.0 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM The Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of Technical Specification 3.12.1. The sampling and analysis program described herein wss developed to provide representative measurements of radiation and radioactive materials resulting from station operation in the principal pathways of exposure of MEMBERS OF THE PUBLIC. This monitoring program implements Section IV.B.2 of Appendix ! to 10 CFR Part 50 and thereby supplements the radiological effluent control program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the affluent measurements and the modeling of the environmental exposure pathways. Guidance for the development of this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring. 6.1 Sampling Locations Sampling locations as required by TS 3.12.1 are described in Table 6-1 and shown on the maps in Figures 6-1, 6-2, 6-3, and 6-4. For purposes of implementing TS 3.12.2, sampling locations will be modified NOTE: as required to reflect the findings of the land use census as described in OOCM Section 5.1. 6.2 Reporting Levels TS 3.12.1, Action b, describes criteria for a Special Report to the NRC if levels of plant-related radioactive material, when everaged over a calendar quarter, exceed the prescribed levels of TS Table 3.12.1-2. The reporting levels are based on the design 3.11.1.2, 3.11.2.2 ob3ctive doses of 10 CFR 50, Appendix 1 (i.e., the annual limits of TS and 3.11.2.3). In other words, levels of radioactive material in the respective sampling medium equal to the prescribed reporting levels are representative of potential annual doses of 3 mrom, total body or 10 mrom, maximum organ from liquid pothways; or 5 mrom, total body, or 15 mrom, maximum organ for the gaseous effluent pathway. These potential doses are modeled on the maximum individual exposure or consumption rates of NRC Regulatory Guide 1.10g. The evaluation of potential doses should be based solely on radioactive material resulting from plant operation. As stated in TS 3.12.1, Action b, the report shall also be submitted if radionuclides other than those in TS Table 3.12.1-2 are detected (and are a result of plant effluents) and the potential dose exceeds the above annual design objectives. The method described in ODCM Section 4.2.3 may be used for assessing the potential dose and required reporting for radionuclides other than those in TS Table 3.12.1-2. ARMS - INFORMATION SERVICES Date approved: Release authorized by: Change numbers incorporated: LCR 88-032-ODM Rev 2' Date DSN DTC TMPLAN File 1715.02 Recipient
P ge 6.0-2 6.3 interteb;rct:ry Comp;rison Pr:gr:m A major objective of this program is to assist laboratories involved in environmental radiation measurements to develop and maintain both an intralsboratory and an interlaboratory quality control program. Th!s is accomplished through an extensive laboratory intercomparison study (* cross-check") program involving environmental media milk, water, sir, food, soll, and gases) and a variety of radionuclides with ( activtties at or near environmental levels. Simuisted environmental semples, containing known sniounts of one or more radionuclides, are prepared and routinely distributed to all laboratories upon request. These laboratories perform the required analyses and return their data to the Quality Assurance Branch of the Environmental Protection A0ency (EPA). Tha EPA performs { statistical analysis and comparison with known values and analytical values obtained j from other participating laboratories. A report and ccatrol chart are retumed to each participant. The program thus enables each laboratory to document the precision and accuracy of its radiation data, identify instrument and procedural problems, and compare its performance with that of other laboratories. The environmental laboratory is required to participate in a Commission-approved Interlsboratory Comparison hogram and to submit GA Program Progress Summary Reports to Detroit Edison on a bimonthly or quarterly basis. These reports contain summary descriptions and performance data summaries on reference standards, blank, blind, spiked, and duplicate analyses, as well as the USEPA and other Laboratory Intercommission Programs, as applicable. A summary of the intertaboratory Comparison Program results obtained is required to be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.g.1.7. Participation in an approved interlsboratory Comparison Program ensures that an independent check on the precision and accuracy of the measurements of radioactive material in environmental sample matrices is performed as part of the QA Program for environmental monitoring in order to demonstrate that the results are valid for the purpose of Section IV.B.2 of Appendix 1 to 10 CFR Part 50. END OF SECTION 6.0
ODCM-6.0 Revision 2 Page 6.0-3 TABLE E.0-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM FERMI 2 SAMPLE LOCATIONS AND ASSOCIATED MEDIA KEY 1-T TLD Locations (Pg. 6.0-4 to 6.0-6) 2-S Sedirnents Locations (Pg. 8.0-7) 3-F Fish Locations (Pg. 6.0-7) 4-M Milk Locations (Pg. 6.0-6) 5-DW Drinking Water Locations (Pg. 6.0-9) 6-SW Suriace Water Locations (Pg. 6.0-9) 7-OW Ground Water Locations (Pg. 6.0-9) 8-API Air Particulate / lodine Locations (Pg. 6.0-10) 9-FP Food Products Locations (Pg. 6.0-11) NOTE: Ferrni 1 sampling information is Appendix 1.
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er F 23 n0 O. e. Gt0o e e Wt e i t d N Cwo o 0cc W 0OP d k h h e l t t t i i t a M W a tcees A d na sne t tec e L e d R lp d m r R a e o S n t b t e s o r 2 t u L o t C p p E w t 1 m i e r r N 5 N 0 o c 0 6 F h s 2 7 9 t e 1 2 3 E D 5 2 t n u 7 3 L9 er m - A e r T e m) o m r rt F r P oo o F r e F m e t n yn u s t t o s e r oC e o e D( R v t t n e s ) a n mo t e or t rp n fp t l t e A m e a m e( n c 2 7 5 o nr r eo 4 5 4 i tt 1 v sce n I E De R l a c t c lt e e et l ce n n ae '9 e I t g9 l l d omt 9 1 1 e I Ac 6 3 3 R e/e 2 / / rrr / w w eoi W N M etD t c ee mS ner t e 2 3 t e - t am # N M t ue St
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3 A a A A u A A 4 21 p s 5 3 g S 5 se @ n C. r .o 8iO F s 3 e 'Civ9 Doc Oc# e t dne s de e w W W w w W w t l S S S u R R a t d c o e B es A er d u n n h e a t t w e t c e S r u t r n k C n t o n d a s a n n t t t a a o w t' e t se e y S k ne f a av g w e. a H s o. t l A d n P t r l t ee e e eR i n e o t e e s d t o t g r ns g a r m u i o s l e l h o t me o sn e er we l O s i s a p . L a tf o f s n m N w af ee oe a u .f t l c C p o e rJ n k T e .c 2 u NA w f e. a n s m W L ro t o o r f ee l t e 0 o r c dt r E t c o at ek w t e t gl f 6 F e n i l b e So R ws t l 1 E a h e u v t m w B e e bP o e rb t y e L ( l e t e n r n b A o w e n t 2 t i l i e a/ i Cl L T o i e o. r i r t n l d e w1 n e p p t b w e. U tB r t eR h t t e i e r e k t os t of t n c r ae a m ru m ro n t s o ee e r no r t e r e h rr r e oe e e. C o D s AC A F Mp F DN r t o t t n c o ) a o e s e R o s e mr 2 t op n rp t t e f a a m t t m ( t l t t n e a m e e e m r o cr 5 e r ne 5 9 4 4 2 4 F 9 t et 1 O 2 m w t c O 1 O O o n ea r E i e f DR s n e e c k t a e h t e l teu e cm '9 '9 l e '8 r '0 6 t l i S 5 'O '9 e d ezn 1 2 t w e eAo 9 5 t t /i 1 / / t 8 / / 1 E E e 7 w E E s rrt eoc 5 N N 5 / 5 M o et e 5 5 5 N C r n t cr e e eei r t mSD e n t a e t o e s o k r t l e g e e O t e e t3 y n e L r r1 a8 t f 1 rt i oe ei er c C oe w t e a h t n C o t o e te f t n cn e yr m yr t es r u a aa k w t n r tt o t u u o w eh a o io o i e N Ss S S S RC L R Cw F CD e
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Nuclear Production - Fermi 2 ODCM-APP-A Off:lto Dese Calculatlin M:nual RevislIn 2 P;ge A-1 APPENDIX A: EVALUATION OF OENERIC CONCENTRATION LIMIT FOR LIQUID EFFLUENTS in accordance with the requiremems of TS 3.3.7.11, the Radioactive Liquid Effluent Monitors. shall be operable with alarm setpoints established to ensure that the concentration of radioactive material at the discharge point does not exceed the MPC value of 10 CFR 20, Appendix B, Table 11. Column 2. The determination of allowable radionuclides concentration and corresponding alarm setpoint is a function of the individual radionuclides distribution and corresponding MPC values. In order to limit the need for routinely having to reestablish the alarm setpoints es a function of changing radionuclides distributions, a generic alarm setpoint can be established. This generic setpoint is based on an evaluation of the anticipated radionuclides distribution in liquid effluents presented in Fermi 2 FSAR, Section 11.2, Table 11.2-9. Based on this distribution, the Concentration Umit CL (or effective MPC) is calculated by the equation: C1 C1 CL = MPC) (A-1) { Where: CL = concentration limit (or effective MPC for the mixture of radionuclides so as not to exceed the unrestricted area MPC of 10 CFR 20 (uCl/ml) concentration of radionuclides i in the mixtura (uCl/ml) Ci = MPCI = the 10 CFR 20, Appendix B, Table !!, Column 2 MPC value for radionucliae I (uCl/ml) Using the above equation and the radionuclides distribution in the Fermi 2 FSAR Table 11.2-6, the value of CL is calculated to be 3.0 E - 06 uCi/ml. Table A-1 presents the data for this calculation. END OF APPENDIX A l ARMS - INFORMATION SERVICES Da'te approved: Release authorized by: LCR 88-032-ODM Change numbers incorporated:_ DSN Rev 2 Date DTC TMPLAN File 1715.02 Recipient
R2visi:n 2 Pcge A-2 TABLE A-1 ) Concentration Limit for Liquid Effluents from Fermi 2 Estimated Annual Nuclide MPC Release
~~Ratio (uCi/ml)~~
(Cl) (Cl/MPCI) Na-24 3E-05 0.0044 1.5E+02 Mn-56 1E-04 0.0097 9.7E+01 Cu-44 2E-04 0.013 6.5E+01 Np-239 1E-04 0.0036 3.6E+01 Y-92 6E-05 0.0027 4.5E+01 1-131 3E-07 0.0022 7.3E+01 I l-132 8E-06 0.011 1.4E+03 1-133 1E-06 0.025 2.5E+04 4-134 2E-05 0.0071 3.6E+02 l-135 4E-06 0.018 4.5E+03 TOTAL 0.097 3.2E+04 CL = [Ci ' = 3.0E-06 uCi/ml ((Cg/MPCg) Radionuclides distribution adapted from Fermi 2 UFSAR, Section 11.2, Table 11.2-9. Radionuclides contributing less than 0.1% to the determination of CL (or the effective MPC) have been excluded. END 3
Nuclear Production - Fermi 2 ODCM-APP-B Off:lts Deso Calculatten Manual Revisi:n 2 rese B-1 c APPENDIX 3: TECHNICAL BASIS FOR EFFECTIVE DOSE PACTORS L80UlO EFFLUENT RELEASES Overview To simplify the dose calculation process, it is conservative to identify a controlling, ) dose-significant radionuclides and to use its dose conversion factor in the dose calculations. Using the total release (i.e., the cumulative activity of all radionuclides) and this single dose conversion factor as inputs to a one-step dose assessment yields a dose calculation method which is both simple and conservative. Formi 2 does not have a large data base of previous releases of radioactive liquid effluents upon which to base the determination of the controlling, dose-significant isotope. The Fermi 2 FSAR, Table 11.2-g presents the estimated annual releases from liquid effluents as calculated using the NRC OALE computer code,(NUREG-0016, Revision 1). Site specific dose conversion factors (Ajo) from ODCM Table 2.0-1 were multiplied by the FSAR estimated annual release quantity to determine a relative dose' significance. Table B-1 presents the results of this relative dose evaluation. Because.Cs-134 is the controlling nuclide for the total body dose and has the highest dose conversion factor among the nuclides evaluated for that dose, the use of its dose conversion factor in the simplified dose assessment method for evaluating the total body dose is demonstrably conservative. Selection of the appropriate dose conversion factor for the maximum organ dose is not so straightforward. inspection of Table 9-1 shows that the thyroid dose is th. controlling organ dose, and it follows that the lodines are the controlling radionuclides. However, this I identification is based upon the FSAR estimate of annual releases. To be adequately conservative when using this simpilfied method, it is appropriate to select the largest dose conversion factor from among all the radionuclides evaluated to assure that offsite doses are not mistakenly underestimated. For the FSAR Table 11.2-g isotopes evaluated, there are a few radionuclides with a higher dose conversion factor than 1-133 for the thyroid dose. Further inspection of Table B-1 shows that P-32 is the major contributor to the dose to the bone, which is the second highest organ dose. P-32 has a high dose conversion factor (1.3g E + 06 mrom/hr por uCi/ml) and would provide additional conservatism if used as the simplifying dose conversion factor. However, analysis for P-32 is not required. P-32 decays by bots emission without any accompanying characteristic gammas. Use of the P-32 dose conversion factor is therefore inappropriate. The next largest dose conversion factor of the evaluated radionuclides is Cs-134 for the dose to the liver at 7.0g E + 05 mrom/hr per uCl/ml. (The dose to the liver is the third largest organ dose.) As Cs-134 is easily measured with gamma spectroscopy, has a long half-life, and a high organ dose conversion factor,it is used as the controlling radionuclides for the simplified maximum organ dose assessment. l l ARMS - INFORMATION SERVICES Date approved: Release authorized by: Change numbers incorporated: LCR 88-032-ODM DSN Rev 2 Date DTC TMPLAN File 1715.02 Recipient l .A
h CDCM-APP-B Revisisn 2 Page B-2 Simplified teethod For evaluation compliance with the dose limits of Technical Specification 3.11.1.2, the following simplified equations may be used: Total Body Dtb = 1.67 E - 32
VOL e A(Ca-134,tb)

C1 DF e Z (B-1) where:
l Dtb = dose to the total body (mram) VOL = volume of liquid effluents released (gel) DF = average circulating water reservoir decent line flow (get/ min) 5, near field dilution factor (derived from Regulatory Guide 1.109) Z = A(Cs-134,tb) = 5.80 E + 05 mrom/hr per uCi/ml, the total body ingestion dose factor for Cs-134 Ci = total concentration of all radionuclides (uCi/ml) 1.67 E - 02 1 hr/60 min = Substituting the value for the Cs-134 total body dose conversion factor, the equation simplifies to: Dtb = 9.69 E + 03
~~VOL e~~
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DF e Z (B-3) where: Dmax = maximum organ dose (mrom) A(Cs-134, liver) = 7.09 E + 05 mrom/hr per UCi/ml, the liver ingestion dose factor for Cs-134
ODCM-APP-B Revisien 2 Pope B-3 Substituting the value for the Cs-134 liver dose conversion factor, the equation simplifies to: ) f Dmax
1.18E + 04

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Ci DF e Z l (B-4) I 1 Tritium is not included in the limited analysis dose assessment for liquid releases, because the potential dose resulting from normal reactor releases is relatively negligible. Furthermore, the release of tritium is a function of operating history and is essentially unrelated to radweste system operations. l l I N____-_-_._---__._______________________o__
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Nuclxr Praducti:n - Fermi 2 ODCM-APP-C Offsite Dese Calculation Manual Revlalon 2 Page C-1 j APPENDIX C: TECHNICAL BASIS FOR EFFECTIVE DOSE FACTORS GASEOUS RADWASTE EFFLUENTS Overview Dose evaluations for releases of gaseous radioactive effluents may be simplified by the use of an effective dose factor rather than radionuclides-specific dose factors. These effective dose factors are applied to the total radioactive release to approximate the various doses in the environment; i.e., the total body, gamma-air, and beta-air doses. The effective dose factors are based on the typical radionuclides distribution in the gaseous radioactive offluents. This approach reduces the analyses to a single multiplication (Keff, Mett, or Neff) times the quantity of radioactive gases released, rather than individual analyses for each radionuclides and summing the results to determine the dose. .Yet the approach provides a reasonable estimate of the actual doses since under normal operating conditions there is relatively little ver!ation in the radionuclides distribution. Determination of Effective Doser Factors Effective dose transfsr factors are calculated by the following equations: (K*f) Keff
~~i l~~
(C-1) where: Keff = the effective total body dose factor due to gamma amtssions from all noble gases released (mrom/yr per uCl/m, effective) Ki = the total body dose factor due to gamma epissions from each noble gas radionuclides i released (mrom/yr per uCi/m, from Table 3-3) fi = the fractional abundance of noble gas radionuclides i relative to the total noble gas activity (L + 1.1 M)eff * ((LI + 1.1 Md e fg) (C-2) where: (L + 1.1 M)eff = the effective skin dose factor due to beta and garrpa emissions from all noble gases released (mrom/yr per uCl/m, effective,) (L) + 1.1 Mj) = the skin dose factor due to beta and gamma emisslores trop each noble gas radionuclides i released (mrom/yr per uCl/m, from Table 3-3) ARMS - N6 FORMATION SERVICES Date approved: Release authorized by: Change numbers incorporated: LCR 88-032-ODM DSN Rev 2 Date DTC TMPLAN File 1715.02 Recipient
Ccvision 2 P;ge C-2 l DM e f, / i i (C-3) where: Moff = the effective air dose factorgue to gamma emissions from all noble gases . released (mrad /yr per uCi/m, effective) = the air dose factor due to gpma emissions from each noble gas radionuclides i Mi released (mrad /yr per uCi/m, from Table 3-3) (N e f ) Nogg = i i (C-4) - where: ) ) j Nett = the effective air dose factorgue to beta emissions from all noble gases released (mrad /yr per uCl/m, effective) j = the air dose factor due to bgte emissions from each noble gas radionuclides i Ni released (mrad /yr per uCl/m, from Table 3-3) .l Normally, past radioactive effluent data would be used for the determination of the effective dose factors. Fermi 2, however, does not have a sufficient operating history at or near full j power to provide.a reasonable data base for determination of the typical radionuclides distribution in gaseous effluents. Therefore, the FSAR estimate of radionuclides concentrations at the site boundary is used as the initial typical distribution. The effective dose factors derived from this distribution are presented in Table C-l. Application To provide en additional degree of conservatism, a factor of 2.0 is introduced into the dose l calculation when the effective dose factor is used. This conservatism provides additional assurance that the evaluation of doses by the use of a single effective dose factor will not significantly underestimate any actual doses in the environment. l For evaluating compliance with the dose limits of Technical Specification 3.11.2.2 the following simplified equations may be used: D = 2.0 e 3.s 7 E - g8
~~X/O e Megg e Oi 7~~
~~(C-5) and 2.0 e 3.17 E - OS~~
~~X/O e Nett e Qi a~~
(C-6) where: D = sir dose due to gamma emissions for the cumulative release of all 7 noble gases (mrad) = sir dose due to beta emissions for the cumulative release of all noble gases (mrad) i t '
ODCM-APP-C R:visisn 2 ~ Page C-3 q i 3 atmospheric dispersion to the controlling site boundary (sec/m ) X/O = 3 Moff = 2.7 E + 03, effective gamma-sir dose factor (mrad /yr per uCi/m ) 3 Neff = 2.3 E + 03, effective beta-air dose factor (mrad /pr por uCI/m ) Og = cumulative release for all noble gas radionuclides (uCl) 3.17 E - 08 = conversion factor (yr/sec) 2.0 = conservatism factor to account for the variability in the effluent data Combining the constants, the dose calculation equations simplify to: D = 1.71 E - 84 e X/O e Qi (C-7) and 1.46 E - 04 e X/O e Qi D = (C-8) The effective dose factors are used for the purpose of facilitating the timely assessment of f radioactive effluent releases, particularly during periods when the computer or l ODCM software may be unavailable to perform a detailed dose assessment. l l l l l l l l I O
p I ODCM-APP-C Revisi:n 2 P ge C-4 1 TABLE C-1 I Effective Dose Factors - Noble Gas Effluents Total Body Skin Dose Samma Air Beta Air Dese Factor Factor Dose Factor Dose Factor Isotope Fractional
~~Kegg (L+1.1Megg)~~
Magg Negg Abundance (mrom/gr per (mrom/gr per (mrad /yg)per (rarad/yg)per uCi/m ) uCi/m ) uCi/m uCi/m Kr-85m 0.10 1.2E+02 2.8E+02 1.2E+02 2.0E+02 Kr-85 0.01 1.6E-01 1.4E+ 01 1.7E-01 2.0E +01 Kr-88 0.04 5.9E+02 7.6E+02 S.1E+02 1.2E+02 Kr-89 0.06 1.0E+03 1.7E+03 1.0E+03 6.4E+02 X -133 0.67 2.0E+02 4.7E+02 2.4E+02 7.0E +02 Xe-135 0.02 3.6E+01 7.9E+01 3.8E+01 4.9E+01 X3-137 0.02 2.8E+01 2.8E+02 3.0E+01 2.5E+02 X3-138 0.07 6.2E* 02 1.0E+03 S.4E+02 3.3E + 02 TOTAL 2.6E+03 4.6E+03 2.7E+03 2.3E+03 Radionuclides distribution as presented 6n ODCM Table 3-1, derwed from Fermi 2 UFSAR, Section 11.3, Table 11.3-5. Kr-90, Kr-91, Xe-139, and Xe-140 have been excluded from the UFSAR distribution because of short half-lives and subsequent decay during environmental transport. Kr-87, Xe-131m, and Xe-133m have been excluded because of their negligible fractional abundance. END I i s 1)!, "1 l \\ / _---___-______-____________D
D. Ralph SyMa Seneur V4e Preudent CH9_troi.t 64% North Disse Htghway r4lO A n Nc* port. u.chegan 4as VIOvl I 43i33 as-4tw March 1, 1989 NRC-89-0022 I U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555
~~References:~~
~~(1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43 (2) Appendix A, Facility Operating License No. NPF-43 Technical Specification 6.9.1.8~~
~~Subject:~~
~~Semi-Annual Radiological Effluent Release Report 's I[ .{ The Semi-Annual Effluent Release Report for Fermi 2 is %/ attached. This report is being transmitted in compliance with Y~~
~~4 Reference 2 and Regulatory Guide 1.21 Revision 1.~~
The attached report covers the period from July 1 through December j 31, 1989. f During this reporting period there were no instances of unmonitored or unplanned radioactive releases from the site. Please direct any questions or requests for additional j information to Joseph Pendergast at (313) 586-1682. \\ m Sincerely, (ny cc: A. B. Davis R. C. Knop W. G. Rogers J. F. Stang Region III}}

NRC-89-0022, Semiannual Radioactive Effluent Release Rept for Jul-Dec 1988

Text

SUMMARY

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1.0 INTRODUCTION

5.0 ASSESSMENT

1.0 INTRODUCTION

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