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Rev 1 to Offsite Dose Calculation Manual. W/One Oversize Encl
	ML18038A167
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Site:	Nine Mile Point
Issue date:	05/31/1986
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TABLE OF CONTENTS SECTION SUBJECT TS SECTION PAG E or TABLE or. TABLE

1.0 INTRODUCTION

2.0 LIQUID EFFLUENTS 2 2.1 Liquid Effluent Monitor Alarm Setpoints 2

. 2. 1.1 Basis 3.11.1.1 2 2.1.2 Setpoint Determination Methodology 3.3.7.10 2

2. 1.2. 1 Liquid Radwaste Effluent Radiation 2-3 Alarm Setpoint

2. 1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculations

2. 1.2.3 Service Water and Cooling Tower Blowdown Effluent Radiation Alarm Setpoint

2. 1.3 Discussion 5 2.1.3.1 Liquid Radwaste Effluent 6-9

2. 1.3.2 Service Water and Cooling Tower Blowdown 10-12 2.2 Liquid Effluent Concentration 3.11.1.1 12 Calculation 4.11.1.1.2 2.3 Liquid Effluent Dose Calculation 3.11.1.2 13-14 3.11.1.3 4.11.1. 2 4.11.1.3- 1 2.4 Liquid Effluent Dose Factor 14-15 Derivation Ait 2.5 Liquid Effluent Sampling 4.11-1 15-16 Representativeness note b 2.6 Liquid Radwaste System Operation 3. 11.1.3 16-17 Table 2-1 Liquid Effluent Detector Response 18 Table 2-2 Ait Liquid Effluent Dose Factor 19 Figure 2-1 Liquid Radwaste Treatment System 3.11.1.3 20-2 7 thru 2-8 Flow Diagrams 3. 11.3 Figure 2-9 Liquid Radiation Monitoring 28 Figure 2-10 Off-line Liquid Monitor 29 3.0 GASEOUS EFFLUENTS 30 3.1 Gaseous Effluent Monitor Alarm Setpoints 30 3.1. 1 Basis 3.11.2. 1 30

3. 1.2 Setpoint Determination Methodology 3 '.7.11 30 3.1.2. 1 Stack Noble Gas Radiation Alarm Setpoint 30-31

3. 1.2.2 Vent Noble Gas Radiation Alarm Setpoint 31-32 3.1.2.3 Offgas Pretreatment Radiation Alarm 32 Setpoint

3. 1.3 Discussion 33 3.1.3.1 Stack Effluent 34

3. 1.3.2 Vent Effluent 35 3.1.3.3 ~

Offgas Process 36 3.2 Gaseous Effluent Dose Rate Calculation 3. 11.2. 1 37 3.2.1 Total Body Dose Rate Due to Noble Gases 3.11.2.1. a 37-38 4.11.2.1.1 3.2.2 Skin Dose Rate Due to Noble Gases 3.11.2.1. a 38-3 9

4. 11.2. 1. 1

-i May 1986 8606110219 860605,

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TABLE OF CONTENTS SECTION SUBJECT TS SECTION PAG E or TABLE or TABLE rr:3 Organ Dose Rate Due to I-1 1, I-1 39~1 Tritium and Particulates with half- 3.11.2.l.b lives greater than 8 days 4.11.2.1.2 3.3 Gaseous Effluent Dose Calculation 3. 11.2. 2 41-42 Methodology 3.11.2.3 3.11.2.5 3.3.1 Gamma Air Dose Due to Noble Gases 3.11.2.2.a , 42

4. 11.2.2 3.3.2 Beta Air Dose Due to Noble Gases 3.11.2.2. b 43 3.3.3 Organ Dose Due to I-131, I-133, Tritium 43-45 and Particulates with half-lives '3.11 2.3 greater than 8 days. 3.11 2.5 4.11.2.3 4.11 .2.5. 1 3.4 Gaseous Effluent Dose Factor Definition 45.

and Derivation 3.4. 1 Bi- Plume Shine Gamma Air Dose Factor 45-47 Vi- Plume Shine Total Body Dose Factor 3.4.2 Ki, Li, Mi and Ni- Immersion Dose Factors 47 3.4.3 Pi- I'odine, Particulate and Tritium 47-51 Organ Dose Rate Factors 3.4.4 Ri- Iodine, Particulate and Tritium 51-5 7 Organ Dose Factors 3.4.5 X/Q and Wv- Dispersion Factors for Dose Rate 58 3.4.6 Ws and Wv- Dispersion Factors for Dose 59 3.5 Gaseous Effluent I-133 Estimation 59 3.6 Use of Concurrent Meteorological Data vs. 59 Historical Data 3.7 Gaseous Radwaste Treatment System 3.11.2.4 59 Operation 3.8 Ventilation Exhaust Treatment System 3. 11.2. 5 60 Operations Table 3-1 Offgas Noble Gas Detector Response 61 Table 3-2 Bi and Vi- Plume Shine Dose Factors 62 Table 3-3 Ki, Li, Mi and Ni- Immersion Dose Factors 63 Table 3-4 Pi- Ground Plane Dose Rate Factors 64 Table 3-5 Pi- Inhalation Dose Rate Factors 65 Table 3-6 Pi- Food (Cow Milk) Dose Rate Factors 66 Table 3-7 Ri- Inhalation Dose Factors for Infant 67-70 to 3-10 Child, Teen and Adult Table 3-11 Ri- Ground Plane Dose Factors 71 Table 3-12 Ri- Cowmilk Ingestion .Dose Factors for 72-75 to 3-15 Infant, Child, Teen and Adult Table 3-16 Ri- Cowmeat Ingestion Dose Factors for 76-7 8 to 3-18 Child, Teen and Adult Table 3-19 Ri- Vegetation Ingestion Dose Factors for 79-81 to 3-21 Child, Teen and Adult

-ii May 1986

TABLE OF CONTENTS S ECTION SUBJECT TS SECTION PAGE or TABLE or TABLE Table 3-22 X/Q, Wv and Ws- Di,spersion Factors for 82 Receptor Locations Figure 3-1 Gaseous Radwaste Treatment System Flow 3.11.2.4 83-85 thru 3-3 Diagrams Figure 3-4 Ventilation Exhaust Treatment System 3.11.2.5 86 Flow Diagrams Figure 3-5 Gaseous Radiation Monitoring 87 Figure 3-6 Gaseous Effluent Monitoring System 88 4.0 URANIUM FUEL CYCLE 3.11.4 89-90 4.1 Evaluation of Doses From Liquid Effluents 4.11.4.1 90-91 4.2 Evaluation of Doses From Gaseous Effluents 4.11.4.1 92 4.3 Evaluation of Doses From Direct Radiation 4.11 .4.2 92 4,4 Doses to Members of the Public Within 6.9. 1.8 93-94 Site Boundary 5.0 ENVIRONMENTAL MONITORING PROGRAM 3.12 95

4. 12 5.1 Sampling Stations 3. 12. 1 95 4.12.1 5.2 Interlaboratory Comparison Program 4. 12.3 95 5.3 Capabilities for Thermoluminescent Dosimeters 97-97 Used for Environmental Measurements Table 5.1 Radiological Environmental Monitoring 3. 12.1 98"100 Program Sampling Locations 4. 12. 1 Table 3.12-1 Note (a) 6.0 DISCUSSION OF TECHNICAL SPECIFICATION 101 REFERENCES 6.1 Table 3.12-1 note g 101 6.2 Table 3.12-1 note h 101 6.3 6.4 Table 3.12-1 note Table 3.12-1 note 1 i 102 102 Figure 5.1-1 Nine Mile Point On-Site Map Figure 5.1-2 Nine Mile Point Off-Site Map Figure 5. 1.3-1 Site Boundaries

-iii May 1986

OFF-SITE DOSE CALCULATION MANUAL (ODCM)

INTRODUCTION This is the OFFSITE DOSE CALCULATION MANUAL (ODCM), referenced in the Nine Mile Point Unit 2 Technical Specification. It describes the methodology for liquid and gaseous effluent monitor alarm setpoint calculations, the methodology for computing the offsite dose due to liquid effluents, gaseous effluents, and the uranium fuel cycle as well as the radiological environmental monitoring and interlaboratory comparison programs.

The ODCM will be reviewed and approved by the NRC. Changes shall be provided in the semi annual radioactive effluent release reports submitted to the NRC.

Section 2 establishes methods used to calculate the Liquid Effluent Monitor Alarm setpoints and to demonstrate compliance with TS Section

3. 11.1.1 limits on concentration of releases to the environment as required in TS Section 3.3.7.10'nd 4.11.1.1.2 respectively.

Additionally, the method used to calculate the cumulative dose contributions from liquid effluents and the methods used to assure thorough mixing and sampling of liquid radioactive waste tanks to be discharged as required in TS Section 4.11.1 .2, 4.11.1.3.1 and Table

4. 11"1 note b respectively are presented.

Section 3 establishes calculational methods used to calculate the Gaseous Effluent Monitor Alarm setpoints and to demonstrate compliance with TS Section 3.11.2.1 limits on dose rates due to gaseous releases to the environment as required in TS Section 3.3.7.11, 4.11.2.1.1 and 4.11 .2.1 .2 respectively. Additionally, the calculational methods used to calculate cumulative dose contributions from gaseous effluents as required in TS Section 4.11 .2.2, 4.11 .2.3 and 4.11.2.5 are presented.

S ection 4 establishes the method used to determine cumulative dose contributions from the Uranium Fuel Cycle as required by TS Section

4. 11.4.1, 4.11.4.2 and 6.9.1.8.

Section 5 establishes the environmental monitoring program as required by TS Section 3.12 and 4,12 including the Interlaboratory Comparison Program required by TS Section 4. 12.3.

Section 6 discusses some of the references contained in TS Table 3.12-1, Radiological Environmental Monitoring Program.

May 1986

0 2.0 LIQUID EFFLUENTS Service Water A and B, Cooling Tower Blowdown and the Liquid Radioactive Waste Discharges comprise the Radioactive Liquid Effluents at Unit 2. (See figure 2-9) Presently there are no temporary outdoor tanks containing radioactive water capable of affecting the nearest known or future water supply in an unrestricted korea. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.

2.1 Liquid Effluent Monitor Alarm Setpoints

2. 1.1 Basis Technical Specification 3. 11.1.1 provide the basis for the alarm setpoints: The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1.3-1) shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases'or dissolved or entrained nobles gases, the concentration shall be limited by 2 x 10-4 microcurie/ml total activity.

2. 1.2 Setpoint Determination Methodology 2.1.2.1 Liquid Radwaste Effluent Radiation Alarm Setpoint This monitors setpoint takes into account the dilution of Radwaste Effluents provided by the Service Water and Cooling Tower Blowdown flows. Detector response for the nuclides to be discharged (cpm) is multiplied by the Actual Dilution Factor (dilution flow/waste stream flow) and divided by the Required Dilution Factor (total fraction of MPC in the waste stream) ~ A safety factor is used to ensure that the limit is never exceeded. Service Water and Cooling Tower Blowdown are normally non-radioactive. If they are found to be contaminated prior to a Liquid Radwaste discharge then an alternative equation is used to take into account the contamination. If they become contaminated during a Radwaste discharge, then the discharge will be immediately terminated and the situation fully assessed.

Normal Radwaste Effluent Alarm Setpoint Calculation'.

Alarm Setpoint < [0.8<<(F/f)*U.(Ci"CFi)]/[H.(Ci/MPCi)) + Background.

Where:

Alarm Setpoint The Radiation Detector Alarm Setpoint, cpm 0.8 Safety Factor, unitless F Nonradioactive dilution flow rate, gpm. Service May 1986

Water Flow ranges from 30,000 to 58,000 gpm.

Blowdown flow is typically 10,200 gpm.

Ci Concentration of isotope tank prior to dilution, uCi/ml i in Radwaste CFi Detector response for isotope i, net cpm/uCi/ml See Table 2-1 for a list of nominal values f The permissible Radwaste Effluent Flow rate, gpm Symbol to denote multiplication.

MP Ci Concentration limit for isotope i from 10CFR20

Background

Appendix B, Table II, Column 2, uCi/ml Detector response when sample chamber is filled with nonradioactive water, cpm Zi(Ci+CFi) The total detector response when exposed to the concentration of nuclides in the Radwaste tank, cpm Zi(Ci/MPCi) The total fraction of the 10CFR20, Appendix B, Table II, Column 2 limit that is in the Radwaste tank, unitless ~ This is also known as the Required Dilution Factor (RDF )

CR*ZiCi An approximation toZi(CiCFi) determined, at each calibration of the effluent monitor, by recording monitor cpm response to a typical radwaste tank mixture analyzed by multichannel analyzer (traceable to NBS) ~ CR is a weighted summation of CF.

F/f An approximation to (F+f)/f, the Actual Dilution Factor in effect during a discharge.

Permissible effluent flow, f, shall be calculated to determine that MPC will not be exceeded in the discharge canal.

f (Dilution Flow) * (1 Fraction Tempering)

(RDF) " 1.5 Fraction Tempering A diversion of some fraction of discharge flow to the intake canal for the purpose of temperature control .

NOTE: If Actual Dilution Factor is set equal to the Required Dilution Factor, then the alarm points required by the above equations correspond to a concentration of 80X of the Radwaste Tank concentration. No discharge could occur, since the monitor would be in alarm as soon as the discharge commenced. To avoid this situation, maximum allowable radwaste discharge flow is calculated using a multiple (usually 1.5 to 2) of the Required Dilution Factor, resulting in discharge canal concentration of 2/3 to 1/2 of MPC.

-3" May 1986

2.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoint Calculation:

The allowable discharge flow rate for a Radwaste tank, when one of the normal dilution streams (Service Water A, Service Water B, or Cooling Tower Blowdown) is contaminated, will be calculated by an iterative process.

Using Radwaste tank concentrations with a nominal radwaste effluent flow rate (200 gpm, for example) the resulting fraction of MPC in the discharge canal will be calculated.

FMPC ~ Zi [Zs(Fs*Cis)/(MPCi*Zs[Fs])]

Then the permissible radwaste effluent flow rate is given by:

f Nominal Flow FMPC"2 The corresponding Alarm Setpoint will then be calculated using the following equation, with f limited as above.

0. 8* Zi (Ci*CFi)

Alarm Setpoint + Background Zi[ Zs(Fs*Cis)/(MPCi*Zs[Fs])]

Where:

Alarm Setpoint ~ The Radiation Detector A1arm Setpoint, cpm 0.8 Safety Factor, Unitless Fs An Effluent flow rate for stream s, gpm Ci Concentration of isotope i in Radwaste tank prior to dilution, uCi/ml Cis Concentration of isotope i in Effluent stream s including the Radwaste Effluent tank undiluted, uCi/m 1 CFi Detector response for istope i, net cpm/nCi/ml See Table 2-1 for a list of nominal values MP Ci Concentration limit for isotope i. from 10CFR20 f

Appendix B, Table II, Column 2, ;iCi/ml The permissible Radwaste Effluent Flow rate, gpm Background Detector response when sample chamber is filled with nonradioactive water, cpm EL( Ci"CFi ) The total detector response when exposed to the concentration of nuclides in the Radwaste tank, cpm Zs [Fs+Cis] The total activity of nuclide i in all Effluent streams, uCi-gpm/ml Zs [Fs] The total Liquid Effluent Flow rate, gpm (Service Water & CT Blowdown & Radwaste)

May 1986

0 2.1.2.3 Service Water and Cooling Tower Blowdown Effluent Alarm Setpoint These monitor setpoints do not take any credit for dilution of each respective effluent stream. Detector response for the distribution of nuclides potentially discharged is divided by the total MPC fraction of the radionuclides potentially in the respective stream. A safety factor is used to ensure that the limit is never exceeded.

Service Water and Cooling Tower Blowdown are normally non-radioactive. If they are found to be contaminated by statistically significant increase in detector response then grab samples will be obtained and analysis meeting the LLD requirements of Table 4.11-1 completed so that an estimate of offsite dose can be made and the situation fully assessed.

Service Water and Cooling Tower Blowdown Alarm Setpoint Equation:

Alarm Setpoint < [0.8*Ei (Ci*CFi)]/[Ei (Ci/MPCi)] + Background.

Where:

Alarm Setpoint The Radiation Detector Alarm Setpoint, cpm 0.8 Safety Pactor, unitless Ci Concentration of isotope i in potential containment, uCi/ml CPi. Detector response for isotope i,, net cpm/uCi/ml See Table 2-1 for a list of nominal values MP Ci Concentration limit for isotope Appendix B, Table II, i from 10CFR20 Column 2, uCi/ml Background ~ Detector response when sample chamber is filled with nonradioactive water, cpm Ei(Ci<<CFi) The total detector response when exposed to the concentration of nuclides in the potential containment, cpm Ei ( Ci/MPCi) The total fraction of the 10CFR20, Appendix B, Table II, Column 2 limit that is in the potential containment, unitless.

CR*EiCi An approximation to Ei(CiCFi) determined, at each calibration of the effluent monitor, by recording monitor cpm response to a typical contaminant mixture analyzed by multichannel analyzer (traceable to NBS). CR is a weighted summation of CPi .

2.1.3 Discussion

-5" May 1986

2.1.3.1 Liquid Radwaste Effluent Monitor The Liquid Radioactive Waste System Tanks are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. At the end of the discharge tunnel in Lake Ontario, a diffuser structure has been installed.

Its purpose is to maintain surface water temperatures low enough to meet thermal pollution limits. However, it also assists in the near field dilution of any activity released'ervice Water and the Cooling Tower Blowdown are also pumped to the discharge tunnel and will provide dilution. If the Service Water or the Cooling Tower Blowdown is found to be contaminated, then its activity will be accounted for when calculating the permissible radwaste effluent flow for a Liquid Radwaste discharge .

The Liquid Radwaste System Monitor provides alarm and automatic termination of release if radiation levels above its alarm setpoint are detected.

The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to gamma and beta radiation. However, because of the metal walls of the sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation. Actual detector response Ei(Ci*CFi), cpm, will be evaluated by placing a sample of typical radioactive waste into the monitor and recording the gross count rate, cpm. A calibration ratio, CR, cpm/uCi/ml, will be developed by dividing the noted detector response, Ei(Ci*CFi) cpm, by total concentration of activity Zi(Ci), uCi/cc.

The quantification of the activity will be completed with gamma spectrometry equipment whose calibration is traceable to NBS. This calibration ratio will be used for subsequent setpoint calculations in the determination of detector response:

Zi. ( Ci*CFi) CR* EL( Ci )

Where the factors are all as defined above.

For the calculation of ZL( Ci/MPCi) the contribution from non gamma emitting nuclides except tritium will be initially estimated based on the expected ratios to quantified nuclides as listed in the FSAR Table 11 .2.5.

Fe-55, Sr-89 and Sr-90 are 2.5, 0.25 and 0.02 times the concentration of Co-60. Periodic analysis of waste for these non gamma emitting nuclides by offsite analysis will provide a better estimate once sufficient activity is present.

Tritium concentration is assumed to equal the latest concentration detected in the monthly Tritium analysis (performed offsite) on liquid radioactive waste tanks discharged or based on the latest tritium detected in the spent fuel pool if liquid radioactive waste tank discharges have not been made within the last 6 months.

6- May 1986

Nominal flow rates of the Liquid Radioactive Waste System Tanks discharged is 165 gpm while dilution flow from the minimum number of Service Water Pumps always in service is over 30,000 gpm, and Cooling Tower Blowdown is 10,200 gpm. Because of the large amount of dilution the alarm setpoint could be substantially greater than that which would correspond to the concentration actually in the tank. Potentially a discharge could continue even if the distribution of nuclides in the tank were substantially different from the grab sample obtained prior to discharge which was used to establish the detector alarm point. To avoid this possibility of "Non representative Sampling" resulting in a erronous assumptions about the discharge of a tank, the tank is recirculated for a minimum of 2.5 tank volumes prior to sampling.

A setpoint of 1.65 time the square root of the background above background will be used until grab sample analysis with the required LLD sensitivity on TS Table 4.11-1 detects activity. The square root of the background is an estimate of the standard deviation of backgrounds 1 .65 times the square root corresponds to approximately the 95% confidence level that the detector response is due to normal variances.

May 1986

A sample calculation is presented below assuming tank concentrations equivalent to the diluted concentration presented in FSAR Table which is the expected concentration of effluent waste after dilution that ll .2.5 are discharged with the design limit for fuel failure (the table below is the undiluted concentration corresponding to a tank 2040 gal per day discharge with only cooling tower blowdown dilution of 10,200 gpm).

ISOTOPE ACTIVITY MPC FRACTION DETECTOR CPM NAME CONCENTRATION OF MPC RESPONSE TOTAL uCi/ml uCi/ml (B/C) cpm/uCi/ml cpm A B C D E F (Ci) (MPCi) ( Ci/MPCi ) (CFi) ( CiCFi)

H3 8.4E-3 3E-3 2.8 NA24 1.7E-6 3E-5 5.7E-2 P32 6.8E-8 2E-5 3.4E-3 CR51 2.0E-6 2E-3 1.03-3 MN54 2.4E-S 1E-4 2.4E-4 8. 42E7 1.98E+0 MN56 3.2E-7 1E-4 3.2E-3 1.2E7 3.9E+1 FE55 3. 5E-7 8E-4 4.3E-4 FE59 1.0E-8 5E-5 2.1E-4 8.63E7 9.0E-1 C058 6. 8E-8 9E-5 7. 6E-4 1.14ES 7.8E+0 C060 1.4E-7 3E-5 4.7E-3 1.65E8 2.4E+1 NI63 3. 5E-10 3E-5 1. 1E-5 NI65 1.8E-9 IE-4 1.8E-5 CU64 4.3E-6 2E-4 2. 1E-2 ZN65 6.8E-8 1E-4 6.8E-4 B R83 3.3E-8 3E-6 1.1E-2 BR84 8.9E-14 1.12ES 1 .OE-5 SR89 3.6E-S 3E-6 1.2E-3 7.8E3 2.8E-4 SR90 2.4E-9 3E-7 7.8E-3 SR91 4. 6E-7 5E-5 9.3E-3 1.22ES 5. 7E+1 SR92 7.6E-S 6E-5 1.2E-3 8.17E7 6. 1E+0 Y91 1.7E-8 3E-5 5.7E-4 2.47E8 4.2E+0 Y92 4.6E-7 6E-5 7.8E-3 2.05E7 9.5 Y93 5. 1E-7 3E-5 1.7E-3 ZR95 2.7E-9 6E-5 4.5E-5 8.35E7 2.3E-1 ZR97 1.0E-9 2E-5 5. 2E-5 NB95 2.7E-9 1E-4 2.7E-5 8.5E7 2.4E-1 M099 6.0E-7 4E-5 1.6E-2 2.32E7 1.4E+1 TC99M 1.2E-6 3E-3 4.1E-4 2.32E7 2.8E+1 RU103 6. 8E-9 SE-5 8. 5E-5 RU105 6.8E-S IE-4 6.3E-4 RU106 1.0E-13 1E-5 1.0E-4 AG110M 3.5E-10 3E-3 1.1E-5 TE129M 1.4E-S 2E-5 7.4E-5 TE131M 2.4E-S 4E-5 6.0E-4 May 1986

"' Vl'h" hh 'f fhfh ',h h 'h jd I ]$ ( h j ' ' h h h hh'hg>>'t

ISOTOPE ACTIVITY MPC FRACTION DETECTOR CPM NAME CONCENTRATION OF MPC RESPONSE TOTAL uCi/ml uci/ml B/C cpm/uci/ml cpm A B C D E F (ci) (MPci) (ci/Mpci) (CFi) ( CicFi )

TE132 2. 9E-9 2E-5 1.5E-4 1.12E8 3. 3E-1 I131 1.4E-6 3E-7 4.7 1 .01E8 1,4E+2 I132 2. 5E-7 8E-6 3.2 2.63E8 6.7E+1 I133 1.2E-5 1E-6 12.3 9.67E7 1.2E+3 I134 7. 2E-10 2E-5 3.6E-5 2.32E8 1.7E-1 I135 3.8E-6 4E-6 9.4E-l 1.17E8 4.4E+2 CS134 5. 1E-6 9E-6 5. 7E-2 1.97E8 1.0E+2 CS136 3.3E-7 6E-5 5.5E-3 2.89E8 9.4E+1 CS137 1.3E-6 2E-5 6. 6E-2 7.32E7 9. 4E-1 CS138 8.4E-12 1.45E8 1.2E-3 BA140 1.3E-7 2E-5 6. 6E-2 4.99E7 6.6E+0 LA142 3.2E-9 3E-6 1.0E+3 CE141 1.0E-8 9E-5 1. 1E-4 CE143 7.6E-9 4E-5 1.9E-4 CE144 7.6E-9 1E-5 1.9E-4 1.03E7 1.0E-2 PR143 1.4E-8 5E-5 2.8E-4 ND147 1.0E-9 6E-5 1.7E-9 W187 6.3E-8 6E-5 1.0E-3 NP239 2.3E<<6 lE-4 2.3E-2

2. 1E+1 2.4E+3 For the example tank, permissible discharge flow to ensure a concentration less than MPC in the discharge canal would be'.

~s 10 200

~

1 cx 324 gpm 2.1El

1.5 Since maximum obtainable Liquid Radwaste discharge flow is 165 gpm, this value would be used for the discharge, and for calculation of the alarm setpoint .

The Liquid Radwaste Effluent Radiation Monitor Alarm Setpoint equation is:

Alarm Setpoint [ 0.8<<F/f* ZL( Ci*CFi))/[H.( Ci/MPci)] + Background .

Where the Alarm Setpoint is in cpm, F is 10,200 gpm, Q.(ci"CFi) is 2.4E+3 cpm, f is 165 gpm and Ef,( Ci/MPci) is 2.1E+1 unitless. These values yield an Alarm Setpoint of 5.7E+3 cpm above background, while the expected detector response is 2.4E+3 cpm. It should be noted that the lack of detector response data for many of the nuclides makes this calculation conservative. Additionally it should be noted that of the tank indicates that no activity detectable above the LLD if grab sample analysis requirements of Table 4.11-1 then the Liquid Radwaste Effluent Radiation Monitor Alarm Setpoint will be set at 1.65*(Background)*"0.5 cpm above Background, cpm.

-9" May 1986

2.1.3.2 Service Water and Cooling Tower Blowdown Effluent Monitor Service Water A and B and the Cooling Tower Blowdown are pumped to the discharge tunnel which in turn flows directly to Lake Ontario. Normal flow rates for each Service Water Pump is 15,000 gpm while that for the Cooling Tower Blowdown is 10,200 gpm. Credit is not taken for any dilution of these individual effluent streams.

The radiation detector is a sodium iodide crystal. It is a scintillation device. The crystal is sensitive to gamma and beta radiation. However, because of the metal walls in its sample chamber and the absorption characteristics of water, the monitor is not particularly sensitive to beta radiation.

Detector response Zi(Ci"CFi) will initially be assumed to correspond to that calculated by the manufacturer. However, this will be evaluated prior to Commercial Operation and during every fuel cycle by placing a diluted sample of Reactor Coolant (after a two hour decay) in the monitor and noting its gross count rate. Reactor Coolant is chosen because represents the most likely contaminate of Station Waters.

A two hour decay is chosen by judgement of the staff of Niagara Mohawk Power Corporation.'Reactor Coolant with no decay contains a considerable amount of very energetic nuclides which would bias the detector response term high. However assuming a longer than 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months decay is not realistic as the most likely release mechanism is a leak through the Residual Heat Removal Heat Exchangers which would contain Reactor Coolant during shutdowns.

The initial setpoint calculation is presented as both an example and for the purposes of documenting the calculation. It will be recalculated prior to commercial operation and during every fuel cycle when a Radiochemical analysis of Reactor Coolant is completed for E bar determination as required by TS Table 4.4.5-1 or when activity is detected in the respective effluent stream.

ISOTOPE 2 HR DECAY MPC FRACTION DETECTOR CPM NAME ACTIVITY OF MPC RESPONSE TOTAL CONCENTRATION uCi/ml B/C cpm/UCi/ml cpm uCi/ml B C D E F (Ci) (MPCi) ( Ci/MPCi) (CFi) ( CiCFi )

H 1.0E-2 3E-3 3.3 F18 1.9E-3 5E-4 3.8 NA24 3.7E-3 3E-5 1.2E-2 P32 7.8E-5 2E-5 3.9 CR51 2.3E-3 2E-3 1.2 MN54 4.0E-5 lE-4 4. OE-1 8.42E7 3.4E3 MN56 2.9E-2 1E-4 2.9E-2 1.2E8 3.5E6 FE55 3.9E-4 8E-4 4. 9E-1 FE59 8.0E-5 5E-5 1.6 8.63E7 6.9E3 C058 5. OE-3 9E-5 5. 6E-1 1.14E8 5. 7E5 C060 5.0E-4 3E-5 1.7E-1 1.65E8 8.3E4 NI63 3.9E-7 3E-5 1.3E-2 May 1986

ISOTOPE 2 HR DECAY MPC FRACTION DETECTOR CPM NAME ACTIVITY OF MPC RESPONSE TOTAL CONCENTRATION <<Ci/ml B/C cpm/ uci/ml cpm

<<Ci/ml A B C D E F (Ci) (MPCi) ( Ci/MPCi) (CFi) ( CicFi )

NI65 3.OE-4 1E-4 3.0 CU64 1. 1E-2 2E-4 5. 5E1 ZN65 7.8E-5 1E-4 7.8E-1 ZN69M 7.4E-4 6E-5 1.2E1 BR83 1.3E-2 3E-6 4.3E3 BR84 2. 1E-3 1.12E8 2.4E5 RB89 1.0E-4 SR89 3. IE-3 3E-6 1.0E3 7.8E3 2.4El SR90 2.3E-4 3E-7 7.7E2 SR91 6. OE-2 5E-5 1 ~ 2E3 1.22E8 7.3E6 SR92 6.6E-2 6E-5 1.1E3 8.17E7 5.4E6 Y91 1.1E-4 3E-5 3.7 2.47ES 2.7E4 Y92 1.3E-2 6E-5 2.2E2 2.05E7 2.7E5 Y93 1.0E-2 3E-5 3.3E2 ZR95 4.OE-5 6E-5 6.7E-1 8.35E7 3.3E3 ZR97 2.9E-5 2E-5 1 ~5 NB95 4. 1E-5 aE-4 4. IE-I 8.5E7 3.5E3 M099 2.2E-2 4E-5 5. 5E-1 2.32E7 5. 1E5 TC99M 2.2E-1 3E-3 7.3E1 2.32E7 5. 1E6 RU103 5.4E-5 8E-5 6.8E-1 RU105 4.5E-3 1E-4 4.5E1 RU106 8. 4E-6 lE-5 8. 4E-1 AG110M 6.0E-5 3E-5 2.0 TE129M 1.1E-4 2E-5 5.5 TE131M 2.7E-4 4E-5 6.8 TE132 4.8E-2 2E-5 2.4E3 1.12ES 5.4E6 I131 lo3E-2 3E-7 4.3E4 1 -01E8 1.3E6 I132 1. 2E-1 SE-6 1.5E4 2.63ES 3.2E7 I133 1.5E-1 1E-6 1.5E5 9.67E7 1.45E7 I134 8.0E-2 2E-5 4.0E3 2.32E8 1.86E7 I135 1.4E-1 4E-6 3.5E4 1.17ES 1.6E7 CS 134 1.6E-4 9E-6 1.8E1 1-97E8 3.2E4 CS136 l. 1E-4 6E-5 1.8 2.89E8 3.2E4 CS137 2.4E-4 2E-5 1.2E1 7.32E7 1.8E4 CS138 1.4E-2 1.45ES 2.0E6 BA140 9.0E-3 2E-5 4.5E2 4.99E7 4.5E5 LA142 7.1E-3 3E-6 2.4E3 CE141 9E-5 CE143 4E-5 CE144 8. 1E-5 lE-5 3.5 1.03E7 3.6E2 PR143 5E-5 ND147 6E-5 F87 6E-5 NP239 2.3E-l lE-4 2.3E3 TOTALS ~ E a.

May, 1986

The Service Water Effluent Radiation Monitor Alarm Setpoint equation is:

Alarm Setpoint ~ [0.8* Ei(Ci*CFi)]/[Ei(Ci/MPCi)]+ Background.

Where the Alarm Setpoint is in cpm, Zi(Ci"CFi) is 1.2E8 cpm, "and Ei(Ci/MPCi) is 2.7E5 unitless. These values yield an Alarm Setpoint of 3.55E2 cpm above background. It should be noted that the lack of detector response data for many of the nuclides makes this calculation conservative.

2.2 Liquid Effluent Concentration Calculation This calculation documents compliance with TS Section 3.11.1.1:

The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1.3-,1) shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10E-4 microcurie/ml total activity.

The concentration of radioactivity from Liquid Radwaste, Service Water A and B and the Cooling Tower Blowdown are included in the calculation. The calculation is performed for a specific period of time. No credit taken for averaging or totaling. The limiting concentration is calculated as follows:

MPC Fraction Ei [ zs(Cis*Fs)/(MPCi" Zs(Fs)) ]

Where.'PC Fraction ~ The limiting concentration of 10 CFR 20, Appendix B, radionuclides Table II, Column 2, for other than dissolved or entrained noble gases. For noble gases, the concentration shall be limited to 2 x 10E-4 microcurie/ml total activity, unitless Ci s The concentration of nuclide i in particular effluent stxeam s, uCi/ml Fs The flow rate of a particular effluent stream s, gpm MPCi ~ The limiting concentration of a specific nuclide i from 10CFR20, Appendix b, Table II> Column 2 (noble gas limit is 2E-4),

ICi/ml Es(Cis<Fs) The total activity rate of nuclide the effluent streams s, uCi/ml *gpm i, in all Xs(Fs) The total flow rate of all effluent streams s, gpm.

A value of less than one for MPC fraction is considered acceptable for compliance with TS Section 3.11.1.1.

-12" May 1986

2.3 Liquid Effluent Dose Calculation Methodology This calculation documents compliance with TS Section 4.11.1.2 and 4.11.1.3.1 for doses due to liquid releases't is completed once per month to assure that TS Section 3.11.1.2 and 3.11.1.3 are not exceeded:

The dos'e or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1.3-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrem to the whole body and to less than or equal to 5 mrem to any organ, and

b. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.

The liquid radwaste treatment system shall be OPERABLE, and appropriate portions of the system shall be used to reduce releases of radioac'tivity when the projected doses due to the liquid effluent, from the unit, to UNRESTRICTED AREAS (see figure 5.1.3-1) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period.

Doses due to Liquid Effluents are calculated monthly for the fish ingestion and drinking water pathways from all detected nuclides in liquid effluents released to the unrestricted areas using the following expression from NUREG 0133, Section 4.3.

Dt EL [Ait*ZL"(dTl*Cil<Fl)]

Where:

Dt The cumulative dose commitment to the total body or any organ, t from the liquid effluents for the total time period E1(dT1),

mrem dT1 The length of the 1 th time period over which Cil and Fl are averaged for all liquid releases, hours Ci1 The average concentration of radionuclide, i, in undiluted liquid effluents during time period dT1 from any liquid release, pci/ml Ait The site related ingestion dose commitment factor to the total body or any organ t for each identified principal gamma or beta emitter, mrem/hr per uCi/ml. Table 2-2.

The near field average dilution factor for Cil during any liquid effluent release Defined as the ratio of the maximum undiluted liquid waste flow during release to the product of the average flow from the site discharge structure to unrestricted receiving waters times 5.9. (5.9 is the site specifice applicable factor for the mixing effect of the discharge structure.) See the Nine Mile Point Unit 2 Environmental Report Operating License Stage, Table 5.4-2 footnote 1.

<3- May 1986

Example Calculation Thyroid A sample of a radwaste tank indicates I-131 and H"3 concentrations of 1.5E-6 and 8.9E-3 uCi/cc respectively. The tank contains 20,000 gallons of waste to be discharged. The tank is discharged at 165 gpm and there is 30,000 gpm of available dilution water:

Dt zi[Ait*E1(dT1*Cil*F1)] ~

Where Dt mrem is the dose to organ t, Ait mrem/hr per gi/ml is the ingestion dose commitment factor, dT hours is the time interval over which the release occurs, Ci uCi/ml is the undiluted concentration of nuclide in the release and Fl unitless is the dilution factor for the release.

i From Table 2-2 Ait is 7.21E4 and 3.37E-1 mrem/hr per pCi/ml respectively for I-131 and H-3 dose to the thyroid. Prom the discharge and dilution flow rate, Pl unitless can be calculated-'l

~ 165gpm/(30,000gpm <<5.9) ~ 9.32E-04.

From the tank volume and discharge rate the length of time required for the discharge is:

dT 20,000 gal/165 gpm ~ 121.2 min ~ 2.02 hr These values will yield 2.04E-4 and 5,65E-6 mrem for I-131 and H-3 respectively for the thyroid when inserted into the equation for Dt. Thus the total dose from the tank is 2.06E-4 mrem to the thyroid. The dose limit to the maximum exposed organ is specified by TS Section 3.11.1.2 and 3.11.1.3.

2.4 Liquid Effluent Dose Factor Derivation Ait Ait mrem/hr per pCi/ml takes into account the dose from ingestion of fish and drinking water. It should be noted that the fish ingestion pathway is the most significant pathway for dose from liquid effluents ~ The water consumption pathway is included for consistancy with NUREG 0133. Drinking water is not routinely sampled as part of the Environmental Monitoring Program because of its insignificance.

The above equation for calculating dose contributions requires the use of dose factor Ait for each nuclide, i, which embodies the dose factors, pathway transfer factors (e.g., bioaccumulation factors), pathway usage factors, and dilution factors for the points of pathway origin. The adult total body and organ dose factor for each radionuclide will be used from Table E-11 of Regulatory Guide 1.109. The dose factor equation for a fresh water site is:

Ait ~ Ko"(Uw/Dw + Uf*BFi)*DFi

->4- May 1986 E' ~ ~/

h I

Where.

Ait Is the composite dose parameter for the total body or organ of an adult for nuclide, i, for all appropriate pathways, mrem/hr per pai/ml Ko Is the unit conversion factor, 1.14E5~1X10E6pCi/ 4i x 1E3 ml/kg -: 8760 hr/hr Uw 730 kg/yr, adult water consumption Uf 21 kg/hr, adult fish consumption BFi Bioaccumulation factor for nuclide, i, in fish, pCi/kg per pCi/1, from Table A-1 of RG 1.109 DFi Dose conversion factor for nuclide, i, for adults in respective organ, t, in mrem/pCi, from Table E-11 of RG 1.109.

Dw Dilution factor from the near field area within one-quarter mile of the release point to the potable water intake for the adult water consumption. This is the Metropolitian Water Board, Onondaga County intake structure located west of the City of Oswego. From the NMP-2 ER-OLS Table 5.4-2 footnote 3 this value is 463.8. However the near field dilution factor, footnote 1 is 5.9. So as to not take double account of the near field dilution the value used for Dw is 463.8/5.9 or 78.6, unitless.

Inserting the usage factors of RG 1.109 as appropriate into the equation gives the following expression:

Ait ~ 1.14E5*(730/Dw + 21*BFi)+DFi.

Example Calculation For I-131 Thyroid Dose Factor for exposure from Liquid Effluents:

DFi ~ 1.95E-3 mRem/pCi BFi ~ 1.5E1 pCi/Kg per pCi/1 UF 21 Kg/yr Dw ~ 78.6 unitless Ko 1.14E5 These values will yield an Ait Factor of 7.21E4 mRem~ per pCi-hr as listed on Table 2-2. It should be noted that only a limited number of nuclides are listed on Table 2 2. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then ODCM ~

it will be calculated and included in a revision to the 2.5 Sampling Representativeness This section covers TS Table 4. 11-1 note b concerning thoroughly mixing each batch of liquid radwaste prior to sampling.

May 1986

Liquid Radwaste Tanks at Nine Mile Point Unit 2 contain a sparger spray ring which assist the mixing o f the tank contents while recirculated prior to sampling. This sparger effectively mixes the tank it is being faster than simple recirculation. Normal recirculation flow is 165 gpm and the tank contains up to 25,000 gallons although the entire contents are not discharged; To assure that the tanks are adequately mixed prior to sampling, it is a plant requirement that the tank be recirculated for the time required to pass 2.5 times the volume of the tank:

Recirculation Time 2.5*T/R Where:

Recirculation Time Is the minimum time to recirculate the Tank, min 2.5 Is the plant requirement, unitless Is the tank volume, gal Is the recirculation flow rate, gpm .

Additionally the Alert Alarm setpoint of the Liquid Radwaste Effluent Radiation Monitor is set at a value corresponding to not more than twice its calculated response to the grab sample. Thus this radiation monitor will alarm if the grab sample is significantly lower in activity than any part of the tank contents being discharged.

Service Water A and B and the Cooling Tower Blowdown are sampled from the radiation monitor on each respective stream. These monitors continuously withdraw a sample and pump it back to the effluent stream.

tubing between the continuously flowing sample and the sample spigot The length of contains less than 200ml which is adequately purged by requiring a purge of at least 1 liter when grabbing a sample.

2.6 Li uid Radwaste S stem 0 eration Technical Specification 3.11.1.3 requires the Liquid Radwaste Treatment System to be OPERABLE and used when projected doses due to liquid radwaste would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31&ay period. Cumulative doses will be determined at least once per 31 days (as indicated in Section 2.3) and doses will also be projected radwaste treatment systems are not being fully utilized.

if the Full utilization will be determined on the basis of utilization of the indicated components of each process stream to process contents of the respective system collection tanks:

1) Low Conductivity (Waste Collector): Radwaste Filter (see Fig. 2-2) and Radwaste Demin. (see Fig. 2-3)

2) High Conductivity (Floor Drains): Floor Drain Filter (see Fig. 2-5) or Waste Evaporator (see"Fig. 2-6)

May 1986

3) Regenerant Waste: Regenerant Evaporator (see Pig. 2-8)

NOTE: Regenerant Evaporator and Was te Evaporator may be use d interchangeably.

The dose projection indicated above will be performed in accordance with the methodology of Section 2.3 when ever Liquid Waste is being discharged without treatment in order to determine that the above dose limits are not exceeded.

May 1986 t e ~ or

TABLE 2-1 LIQUID EFFLUENT DETECTORS RESPONSES

NUCLIDE (CPM/ pCi/ml) x 10 Sr 89 0.78E>>04 Sr 91 1.22 Sr 92 0. 817 Y 91 2.47 Y 92 0. 205 Zr 95 0.835'.

Nb 95 85 Mo 99 0.232 Tc 99m 0. 232 Te 132 1.12 Ba 140 0.499 Ce 144 0.103 Br 84 l. 12 I 131 1.01 I 132 2. 63 I 133 0.967 I 134 2. 32 I 135 1.17 .

Cs 134 1. 97 Cs 136 2.89 Cs 137 0. 732 Cs 138 1.45 Mn 54 0. 842 Mn 56 1.2 Fe 59 0. 863 Co 58 l. 14 Co 60 1. 65

Values from SWEC purchase specification NMP2-P281F .

May 1986

TABLE 2-2 Ai q VALUES - LIQUID+

mrem - ml hr uCi NU GLIDE T BODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H 3 3.37E-1 3.37E-1 3.37E-1 3.37E-1 3.37E-1 3.37E-1 Cr 51 1.28 3.21E2 2. 81E-1 7. 63E-1 1. 69 Mn 54 8.36E2 1.34E4 4.38E3 1.30E3 Fe 59 9.40E2 8. 18E3 1.04E3 2.45E3 6.85E2 Co 58 2.01E2 1.82E3 9.00E1 Co 60 5.70E2 4.85E3 2.58E2 Zn 65 3.33E4 4.65E4 2'.32E4 7.38E4 4.93E4 Sr 89 6.44E2 3.60E3 2.24E4 Sr 90 1.36E5 1.60E4 5.52E5 Zr 95 5.91E-2 2.77E2 2.72E-1 8.74E-2 1.37E-1 Mo 99 2.05El 2.50E2 1.08E2 2.44E2 I 131 1.26E2 5.80E1 1. 54E2 2. 20E2 3. 7 7E2 7. 21E4 I 133 2.78E1 8.21E1 5.25E1 9. 13E1 1.59E2 1.34E4 Cs 134 5.79E5 1.24E4 2. 98E5 7. 09E5 2. 29E5 7. 61E4 Cs 136 8.86E4 1.40E4 3. 12E4 1.23E5 6.85E4 9.39E3 Cs 137 3.42E5 1.01E4 3.82E5 5.22E5 1-77E5 5.89E4 Ba 140 1.41El 4.45E2 2.16E2 2.71E-l 9.22E-2 1.57E-1 Ce 141 2.48E-3 8.36El 3.23E-2 2.19E-2 1.02E-2 Nb 95 1.34E2 1.51E6 4.47E2 2.49E2 2.46E2 La 140 2.03E-2 5.63E3 1.52E-1 7.67E-2 Ce 144 9.05E-2 5.70E2 1.69 7.04E-1 4. 18E-1

Calculated in accordance with NUREG 0133, Section 4.3.1 C9- May 1986

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INTERCHANGEABLE PURGE OR TEST CONNECTION GRAB S SAMPLER I (2) I TI I I TS I (2) I S CS DATA LIDUID ACQUISITION DETECTOR UNIT L (DAU) SAMPI.ER I I (8) FI FS Pl PUMP NOTES: (1) GLOBE VALVE, ALL OTHER MANUALLY (6) TS-TEMPERATURE SWITCH OPERATED VALVES ARE BALL VALVES (7) CS-CHECK SOURCE (2) REOUIRED ONLY IF SAMPLE FLUID TEMPERATURE EXCEEDS SELLERS (8) PI-PRESSURE INDICATOR DETECTOR TEMPERATURE REOUIREMENTS (9) FI-FLOW INDICATOR (3) ~ NORMALLYCLOSED (4) fOCg NORMALLYOPEN (10) FS-FLOW SWITCH (11) DRAIN CONNECTION (5) TI-TEMPERATURE INDICATION ODCM Fig. 2-10 FIGURE 11.5-3 OFF-LINE LIQUID MONITOR NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT May 1986
3.0 GASEOUS EFFLUENTS The gaseous effluent release points are the stack and the combined Radwaste/Reactor Building vent. (See Figure 3.5) The stack effluent point includes Turbine Building ventilation, main condenser offgas (after charcoal bed holdup), and Standby Gas Treatment System exhaust. NUREG 0133 and Regulatory Guide 1.109, Rev. 1 were followed in the development of this section.
3. 1 Gaseous Effluents Monitor Alarm Setpoints 3.1.1 Basis Technical ,Specification Section 3.11.2. 1 and 3.11.2.7 provide the basis for the gaseous effluent monitor alarm setpoints.
TS Section 3.11.2.1: The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDRY (see Figure 5. 1.3-1) shall be limited to the following:
a. For noble gases: Less than or equal to 500 mrem/yr to'he whole body and less than or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, for iodine-133, for tritium, and for all radionuclides with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ.
TS Section 3.11.2.7: The radioactivity rate of noble gases measured downstream of the recombiner shall be limited to less than or equal '350,000 microcurie.es/second during offgas system operation. 3.1.2 Setpoint Determination Methodology The alarm setpoint for Gaseous Effluent Noble Gas Monitors are based on a dose rate limit of 500 mrem/yr to the Whole Body. These monitors are sensitive to only noble gases. Because of this considered impractical to base their alarm setpoints on organ dose it is rates due to iodines or particulates. Additionally skin dose rate is never significantly greater than the whole body dose rate. The alarm setpoint for the Offgas Noble Gas monitor is based on a limit of 350 000 uCi/sec. This is the release rate for which a FSAR accident analysis was completed. At this rate the Offgas System charcoal beds will not contain enough activity so that their failure and subsequent release of activity will present a significant offsite dose assuming accident meterology.
3. 1.2. 1 Stack Noble Gas Detector Alarm Setpoint Equation:
0.&R* ZL(Ci) Alarm Setpelat < ~EE Ci~epl Alarm Setpoint Is the alarm setpoint of the Stack Effluent Monitor, uCi/se c
-30" May 1986
0.8 Is a Safety Factor, uni ties s Is a value of 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total rate corresponds to <500 mrem/yr Is the concentration of nuclide i, uCi/ml Is the Stack effluent flow rate, ml/sec Is the constant for each identified noble gas nuclide accounting for the whole body dose from the elevated finite plume listed on Table 3-2, mrem/yr per uCi/sec Is the total concentration of noble gas nuclides in the Stack effluent, uCi/ml a.(ci*Vi) Is the total of the product of the each isotope concentration times its respective whole body plume constant, mrem/yr per ml/sec. It should be noted that the flow rate of the Stack effluent has been canceled out of the above expression. The equation ratios the basis, R, to the actual dose rate from the effluent, F*Ei (Ci<Vi), and multiplies the unitless result by the actual effluent release rate, F*EL(Ci). Since the Stack Effluent Monitor actually measures release rate in uCi/sec the detector response does not enter in. 3.1 .2.2 Vent Noble Gas Detector Alarm Setpoint Equation: 0.8*R*Ei (Ci) Alarm Setpoint < Where: Alarm Setpoint Is the alarm setpoint of the Vent Effluent Monitor, uCi/sec 0.8 Is a Safety Factor Is a value of 500 mrem/yr or less depending upon the dose rate from other release points within the site such that the total rate corresponds to < 500 mrem/yr Ci Is the concentration of nuclide i, uCi/ml Is the Vent effluent flow rate, ml/sec (X/Q)v Is the highest annual average atmospheric dispersion coefficient at the site boundry as listed in the Final Environmental Statement, NUREG 1085, Table D-2, 2.0E-6 se c/m3 Is the constant for each identified noble gas nuclide accounting for the whole body dose from the semi-infinite cloud listed on Table 3-3, mrem/yr per uCi/m3 May 1986
zL(ci) Is the total concentration of noble gas nuclides in the Vent effluent, uci/ml Zi(ci<<Ki) Is the total of the product of the each isotope concentration times its respective whole body immersion constant, mrem/yr per ml/m3 It should be noted that the flow rate of the Vent effluent has been canceled out of the above expression. The equation ratios the basis, R, to the actual dose rate from the effluent, F*(X/Q)v*Ei(ci*Ki) and multiplies the unitless result by the actual effluent release rate, F* Zi.(ci). Since the Vent Effluent Monitor actually measures release rate in uci/sec the detector response does not enter in. 3.1 .2.3 Offgas Pretreatment Noble Gas Detector Alarm Setpoint Equation: 0.8*350,000*2.1E-3*Zi(ci*CFi) Alarm Setpoint < f Ei Ci + Background Where.'larm Setpoint Is the alarm setpoint for the offgas pretreatment Noble Gas Detector, cpm 0.8 Is a Safety Factor, unitless 350,000 Is the Technical Specification Limit for Offgas Pretreatment, uci/sec
2. 1E-3 Is a unit conversion, 60 sec/min / 28317 ml/CF Ci Is the concentration o f nuclide, i, in the Offgas, uci/ml CFi Is the Detector response to nuclide cpm/uci/ml See Table 3-1 for a list of nominal i, net values. See section 3.1.3.3 for discussion Is the Offgas System Flow rate, CFM Background Is the detector response when its chamber is filled with nonradioactive air, cpm Zi(cicFi) Is the summation of the product of the nuclide concentration and corresponding detector response, net cpm Zi(Ci) Is the summation of the concentration of nuclides in offgas, uci/ml May 1986
Discussion The Stack at Nine Mile Point Unit 2 receives the Offgas after charcoal bed delay, Turbine building ventilation and the Standby Gas Treatment system exhaust. The Standby Gas Treatment system exhaust the primary containment during normal shutdowns and maintains a negative pressure on the Reactor Building during secondary containment isolation. The Standby Gas Treatment will isolate on high radiation during primary containment purges. The Stack is considered an elevated release because its height (131m) is more than 2.5 times the height of any adjacent buildings. Nominal flow rate for the stack is 102,000 CFM. The Offgas system has a radiation detector downstream of the recombiners and before the charcoal decay beds'he offgas, after decay, is exhausted to the main stack. The system will automatically isolate radiation ifabove its pretreatment radiation the alarm setpoint. monitor detects levels of The Vent contains the Reactor Building ventilation above and below the refuel floor and the Radwaste Building ventilation effluents. The Reactor Building Ventilation will isolate when radiation monitors detect high levels of radiation (these are seperate monitors, not otherwise discussed in the ODCM). It is considered a combined elevated/ground level release because even though any adjacent buildings it it is higher than is not more than 2.5 times the height. Nominal flow rate for the vent is 237,310 CFM. Nine Mile Point Unit 1 and the James A Fitzpatrick nuclear plants occupy the same site as Nine Mile Point Unit 2. Because of the independance of these plants safety systems, control rooms and operating staffs it is assumed that simultaneous accidents are not likely to occur at the different units. However, there are two release points at Unit 2. It is assumed that if an accident were to occur at Unit 2 that both release points could be involved'hus the factor R which is the basis for the alarm setpoint calculation is nominally taken as equal to 250 mRem/yr. If there are significant releases from any gaseous release point on the site (>25mRem/yr) for an extended period of time then the setpoint will be recalculated with an appropriately smaller value for R. Initially, and in accordance with Specification 4.3.7. 11, the Germanium multichannel analysis systems of the Stack and Vent will be calibrated with gas, or with cartridge standards (traceable to NBS) in accordance with Table 4.3.7.11-1, note (c) The quarterly Channel
~
Functional Test will include operability of the 30cc chamber and the dilution stages to confirm monitor high range capability. (See Figure 3-6). May 1986
3.1.3. 1 Stack Noble Gas Detector Alarm Setpoint This detector is made of germanium. It is sensitive to only gamma radiation. However, because it is a computer based multichannel analysis system it is able to acurately quantify the activity released in terms of iCi of specific nuclides. Only pure alpha and beta emitters are not detectable, of which there are no common noble gases. A distribution of Noble Gases corresponding to offgas is chosen for the nominal alarm setpoint calculation. Offgas is chosen because it represents contaminate of gaseous activity in the plant.
, the most significant The following calculation will be used for the initial Alarm Setpoint ~ It will be recalculated if a significant release is encountered.
the actual distribution of noble gases will be used in the In that case calculation. The listed activity concentrations Ci, correspond to offgas concentration expected with the plant design limit for fuel failure . ISOTOPE ACTIVITY PLUME PLUME NAME CONCENTRATION FACTOR FACTOR uCi/ml mrem-sec mrem/ r yr-' uCi ml sec B D (B*C) (Ci) (Vi) (Ci~Vi) KR83 8.74E-2 KR85 4.90E-4 3.28E-5 1. 61E-8 KR85M '1.56E-1 3.21E-3 5.01E-3 KR87 5. 23E-1 9.98E-3 5. 22E-3 KR88 5.32E-1 2. 21E-2 l. 18E-2 KR89 1.63 l. 92E-2 3.13E-2 KR90 1.51E-2 XE131M 3. 82E-4 6.55E-5 2.50E-8
'XE133 2.06E-1 5.93E-4 1.22E-4 XE133M 7.35E-3 3.44E-4 2.53E-6 XE135 5. 88E-1 6.12E-3 3.60E-3 XE135M 5. 91E-1 6.12E-3 3.62E-3 XE137 2. 11 2.88E-3 6.08E-3 XE138 1.93 1.33E-2 2.57E-2 AR41 1.61E-2 TO ALS 8. 9. 28E-2 The alarm setpoint equation is:
Alarm Setpoint ~ 0.8*R+Zi(Ci)/Zi(Ci+Vi). Where the Alarm Setpoint is in uCi/sec, R is taken as 250mrem/yr, Z(Ci) is 8.36 uCi/ml and Z(Ci*Vi) is 9.28E-2 mrem/yr per ml/sec. These values yield an alarm setpoint of 1.80E4 pCi/sec. May 1986
3.1.3.2 Vent Effluent Noble Gas Detector Alarm Setpoint This detector is made of germanium. It is sensitive to only gamma radiation. However, it because it is a computer based multichannel analysis system is able to accurately quantify the activity released in terms of uCi of specific nuclides. Only pure alpha and beta emitters are not detectable, of which there are no common noble gases. A distribution of Noble Gases corresponding to that expected with the design limit for fuel failure offgas is chosen for the nominal alarm setpoint calculation. Offgas is chosen because represents the most significant contaminate of gaseous activity in it the plant. The following calculation will be used for the initial Alarm Setpoint. It will be recalculated encountered. if a significant release is In that case the actual distribution of noble gases will be used in the calculation. ISOTOPE ACTIVITY IMMERSION IMMERSIO N NAME CONCENTRATION FACTOR FACTOR uCi/ml mre m~3 mr em~3 A yr-uCi C y~ D~(B*C) KR83 8.74E-2 7. 56E-2 6. 63E-3 KR85 4.90E-4 1.61E-l 7.90E-3 KR85M KR87 1.56E-1 5.23E-1 1.17E-3 l. 82E2 5.92E3 3. 10E3 KR88 5. 32E-1 1.47E4 7.82E3 KR89 1.63 1.66E4 2.71E4 KR90 1.56E4 XE131M 3.82E-4 9. 15El 3.50E-2 XE133 2.06E-1 2. 94E2 6. 06E1 XE133M 7.35E-3 2.51E2 1.84 XE135 5. 88E-1 1. 81E3 1.06E3 XE135M ~ 5.91E-1 3, 12E3 1.84E3 XE137 2. 11 1.42E3 3.00E3 XE138 1.93 1.83E3 1.70E4 AR41 8.84E3 TOTALS 8.36 6. 12E4 The Vent Effluent Noble Gas Monitor Alarm Setpoint equation is'. Alarm Setpoint 0.&R*H(Ci)/[(X/Q)v*ZHCi"Ki)] ~ Where, the Alarm Setpoint is in uCi/sec, R is 250mrem/yr, H.(Ci) is 8.36 uCi/ml, (X/Q) is 2.0E-6 sec/m3 and Zi (Ci "Ki) is 6. 12E4 mrem/yr per ml/m3. This will yield an alarm setpoint of 1.41E4 uCi/sec. May 1986
3. 1.3.3 Offgas Noble Gas Detector Alarm Setpoint The Radiation Detector is a sodium iodide crystal. It is a scintillation device and has a thin mylar window so that sensitive to both gamma and beta radiation. Detector response it is Ei(Ci"CFi) will be evaluated from isotopic analysis of offgas analyzed on a multichannel analyzer, traceable to NBS, prior to commercial operation. A distribution of offgas corresponding to that expected with the design limit for fuel failure is used to establish setpoint initially, assuming the nominal response listed on Table 3-1. The monitor nominal response values will be confirmed during initial calibration using a Transfer Standard source traceable to,the primary calibration performed by the vendor. However, a revision to the ODCM will contain an updated distribution and total detector response based on actual plant experiences. The initial calculation is presented below.
ISOTOPE ACTIVITY DETECTOR DETECTOR NAME CONCENTRATION RESPONSE CPM uCi/ml cpm/uci/ml cpm A B C D (Ci) (CFi) ( Ci*CFi ) KR83 8.74E-2 KR85 4.90E-4 4.30E3 2.11 KR85M 1.56E-1 4.80E3 7.50E3 KR87 5. 23E-1 8.00E3 4.18E3 KR88 5. 32E-1 7.60E3 4.04E3 KR89 1.63 KR90 XE131M 3.82E-4 XE133 2.06E-1 1.75E3 3.60E2 XE133M 7. 35E-3 XE135 5.88E-1 5. 10E3 3.00E3 XE135M 5. 91E-1 XE137 2. 11 8.10E3 1.73E4 XE138 1.93 7. 10E3 1.37E4 AR41
~ 99E The Offgas Noble Gas Monitor Alarm Setpoint equation is:
Alarm Setpoint ~ 0.8*350,000+2.1E-3*Xi(Ci*CFi)/[f"Ei(Ci)]+ Bkg. Where the Alarm Setpoint is in cpm, Zi(Ci*CFi) is 4.99E4 cpm, f is 25CFM and ZL(Ci) is 8.36 uCi/cc. This will yield an alarm setpoint of 1.40E5 cpm above background. Particulates and Iodines are not included in this calculation because this is a noble gas monitor. May 1986
3.2 Gaseous Effluents Dose Rate Calculation This section covers TS Section 4.11.2.1.1 and 4.11.2.1.2 concerning the calculation of dose rate from gaseous effluents for compliance with TS Section 3. 11.2.1. TS Section 3.11.2.1: The dose rate from radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the following:
a. For noble gases: Less than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mrem/yr to the skin, and

b. For iodine-131, iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem/yr to any organ:
3.2. 1 Whole Body Dose Rate Due to Noble Gases This calculation covers TS Section 3.11 .2.1.a (for whole body) and
4. 11.2. 1.1. The dose from the plume shine of elevated releases is taken into account with the factor Vi. The dose from Vent releases takes into account the exposure from immersion in the semi-infinite cloud and the dispersion from the point of release to the receptor which is at the East site boundary. The release rate is averaged over the period of concern. The factors are discussed in greater detail later.
Whole body dose rate due to noble gases: mrem/yr Zi [Vi*Qis + Ki (X/Q)v*Qiv ] Where: Vi Is the constant accounting for the gamma radiation from the elevated finite plume of the Stack releases for each identified noble gas nuclide, i. Listed on Table 3-2, mrem/yr per uCi/sec Qis Is the release rate of each noble gas nuclide, Stack release averaged over the time period i, from the of concern, uCi/se c Ki is >>the constant accounting for the whole body dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table 3-3, mrem/yr per uCi/m3 37 May 1986
(X/Q)v Is the highest calculated annual average relative concentration at or beyond the site boundry for the Vent. Final Environmental Statement, NUREG 1085, Table D-2, 2.0E-6 se c/m3 Qiv Is the release rate of each noble Vent release averaged gas nuclide, i, from the over the time period of concern, uCi/se c Example Calculation: Assume an analysis of the Stack and Vent Effluents indicate that 1.81E4 and 1.26E4 uCi/sec of Xe-133 are being released from each point respectively. From Table 3-2, Vi is 5.93E-4 mrem/yr per uCi/sec. From Table 3-3 Ki is 2.94E2 mrem/yr per pCi/m3. (X/Q)v is 2.0E-6 sec/m3. These values yield a whole body dose rate of 10.7 and 7.41 mrem/yr from the Stack and Vent respectively for a total of 18.1 mrem/yr. This value is added to the whole body dose rates obtained from the Nine Mile Point-Unit 1 and James A. Fitzpatrick plants to obtain the site dose rate to the whole body from noble gas releases. The whole body dose rate due to noble gases is specified by TS Section 3.11.2.l.a. Skin Dose Rate Due to Noble Gases This calculation covers TS 'Section 3. 11.2.1.a ( for skin) and 4.11.2.1.1. For Stack releases this calculation takes into account the exposure from beta radiation of a semi infinite cloud by use of the factor Li. Additionally the dispersion of the released activity from the stack to the receptor is taken into account by use of the factor (X/Q) . Gamma radiation exposure from the overhead plume is taken into account by the factor 1.1Bi. For vent releases the calculations also take into account the exposure from the beta and gamma radiation of the semi infinate cloud by use of the factors Li and 1.1Mi respectively. Dispersion is taken into account by use of the factor (X/Q). The release rate is averaged over the period of concern. The factors are discussed in greater detail later. Skin dose rate due to noble gases
~ Ei [ (Li*(X/Q)s + 1.1*Bi)<<Qis + .'rem/yr (Li + 1.1*Mi)*(X/Q)v+Qiv]
Where: Is the constant to take into account the skin dose due to each noble gas nuclide, i, from immersion in the semi-infinite cloud, mrem/yr per pCi/m3 Is the constant accounting for the air gamma dose rate from immersion in the semi-infinite cloud for each identified noble gas nuclide, i. Listed on Table 3-3, mrad/yr per uCi/m3 1.1 is a unit conversion constant, mrem/rad May 1986
Bi Is the constant accounting for the air gamma dose rate from exposure to the overhead plume o f elevated releases of each identified noble gas nuclide, Listed on Table 3-2, mrad/yr per pCi/sec. (X/Q)v Is the highest calculated annual average relative concentration at or beyond the site boundary for the Vent. Final Environmental Statement, NUREG 1085, Table D-2, 2.0E-6 sec/m3 (X/Q) s Is the highest calculated annual average relative concentration at or beyond the site boundary for the Stack. Final Environyental Statement, NUREG 1085, Table D-2, 4.5E-8 sec/m~ Qi v Is the release rate of each noble gas nuclide, the Vent release averaged over the time period of i, from concern, uCi/se c Qis Is the release rate of each noble gas nuclide, i, the Stack release averaged over the time period of from concern, uCi/se c Example Calculation: Assume an analysis of the Stack and Vent Effluents indicate that 1.81E4 and 1.26E4 uCi of Xe-133 are released from each point. From Table 3-2, Bi is 6.12E-4 mrad/yr per uCi/sec. From Table 3-3, Li and Mi are 3.06E2 and 3.53E2 mrem.mrad/yr per pCi/m3 respectively . (X/Q) for the Stack and Vent is 4.5E-8 and 2.0E-6 sec/m3 respectively. These values yield a skin dose rate of 12.6 and 17.5 mrem/yr for the Stack and Vent respectively for a total rate of 30.1 mrem/yr. This value is added to the skin dose rates obtained from Nine Mile Point-Unit 1 and the James A. Fitzpatrick plants to obtain the site dose rate to the skin from noble gas releases. The skin dose rate limit due to noble gases is specified by TS Section 3.11.2.1.a. 3.2.3 Organ Dose Rate Due to I-131, I-133, Tritium, and Particulates with Half-lives greater than 8 days. This calculation covers TS Section 3.11.2.1.b and 4.11.2.1.2. The factor Pi takes into account the dose rate received from the ground plane, inhalation and food (cow milk) pathways. Ws and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. The release rate is averaged over the period of concern. The factors are discussed in greatez detail later. May 1986
Organ dose rates due to iodine-131, iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days: mrem/yr Zp [ Zi Pip [WsQis + WvQiv] j Where.'ip Is the factor that takes into account the dose to an individual organ from nuclide i through pathway p. For inhalation pathway, mrem/yr per uCi/m ~ For ground and food pathways, m mrem/yr per Ri/sec . Zi Is the summation over all nuclides, i Zp Is the summation over all pathways Ws, Wv Are the dispersion parameters for stack and vent re ease resyectively for each pathway as approriate sec/m or 1/m . See Table 3-22. Qis, Qiv Are the release rates for nuclide vent respectively pCi/sec. i, from the stack and Example Calculation Assume an analysis of the Stack and Vent Effluents indicate that 1.84E-1 and 1.26E-l uCi/sec of I-131 are released .from each point respectively. From Table 3-4 thru 3-6 and 3-22 the following table can be made: ORGAN Pi GROUND Pi INHALATION Pi FOOD or m2~rem/yr mrem/yr m2~re m/yr FACTOR iiCi/sec pci/m3 uCi/sec T BODY 2.46E7 1.96E4 1.43E9 SKIN 2.98E7 BONE 3.79E4 2.77E9 LIVER 4.44E4 3. 26E9 THYROID 1.48E7 1.07E12 KIDNEY 5. 18E4 3. 81E9 LUNG G I-LLI 1.06E3 1.16E8 Ws 1.34E-9 8.48E-9 3.64E-10 Wv 2.90E-9 1.42E-7 4.73E-10 WsQs+WvQv 6.12E-10 1.95E-8 1.27E-10 NOTE: The Dispersion Parameters given in Table 3-22 will be revised based on the results of environmental surveys and meteorological data . From these values the following table of dose rates (mrem/yr) can be calculated: May 1986
ORGAN GROUND INHALATION FOOD TOTAL T BODY 1.51E-2 3.82E-4 1.82E-1 1. 97E-1 SKIN 1.82E-2 1 .82E-2 BONE 7.39E-4 3. 5 2E-1 3. 53E-1 LIVER 8.66E-4 4. 14E-1 4.15E-1 THYROID 2. 89E-1 1.36E+2 1.36E+2 KIDNEY 1.01E-3 4.84E-1 4.85E-1 LUNG GI-LLI 2.07E-5 1.47E-2 1.47E-2 In this case the maximum dose rate to an organ is 136 mrem/yr to the thyroid from I-131. This calculation would be repeated for all nuclides and age groups then summed for each age group to obtain 'the dose rates to all organs. The dose rate limit to the maximum exposed organ is specified by TS Section 3.11.2.l.b. 3.3 Gaseous Effluent Dose Calculation Methodology TS Section 3.11.2.2: The air dose from noble gases released in gaseous effluents, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure 5.1.3-1) shall be limited to the following.
a. During any calendar quarter: Less than or equal to 5 mrad for gamma radiation and less than or equal to 10 mrad for beta radiation, and

b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.
May 1986
TS Section 3.11.2.3: The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium, and all radioactive material in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure
5. 1.3-1) shall be limited to the following:

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

b. During any calendar year: Less than or equal to 15 mrem to any organ.
TS Section 3.11.2.5: The VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE and appropriate portions of this system shall be used to reduce releases of radioactivity when the projected doses in 31 days from iodine and particulate releases, from each unit, to areas at or beyond the SITE BOUNDARY (see Figure 5.1.3-1) would exceed 0.3 mrem to any organ of a MEMBER OF THE PUBLIC. Gamma Air Dose Due to Noble Gases This calculation covers TS Section 3.11.2.2 and 4.11.2.2. Gamma air dose due to noble gases released is calculated monthly. The factor Mi takes into account the dose from immersion in the semi-infinite cloud of the vent release. The factor X/Q takes into account the dispersion of vent releases to the most conservative location. The factor Bi takes into account the dose from exposure to the plume of the stack releases. The release activity is totaled over the period of concern. The factors are discussed in greater detail later. Gamma air dose due to noble gases.' Zj. ~Mi{X/Q)v Qiv + Bi Qis] Where the constants have all been previously defined. Note tha t since Q is expressed as uCi/sec, the constant 3.17E-8 sec given in NUREG-0133, section 5.3. 1 may be omitted, provided that the annual dose calculated is divided by 4 to yield quarter dose, or 12 to yield monthly dose, as applicable. Example Calculation Assume an analysis of the Stack and Vent Effluents indicate that 1.42E11 and 9.91E10 uCi of Xe-133 are released from each point respectively over the last quarter. This correlates to 1.81E4 and 1.26E4 pCi/sec respectively. From Table 3-2, Bi is 6.12E-4 mrad/yr per uCi/sec. From Table 3-3 Mi is 3.53E2 mrad/yr per pCi/m3. (X/Q)v is 2.0E-6 sec/m3. These values yield a gamma air dose rate of 11.1 and 8.9 mrad/yr from the Stack and Vent respectively for a total of 20.0 mrad/yr or 5.0 mrad for the quarter. The gamma air dose limit due to noble gases is specified by TS Section 3.11.2.2. May 1986
Beta Air Dose Due to Noble Gases This calculation covers TS Section 3.11.2.2 and 4.11 .2.2. Beta air dose due to noble gases released is calculated monthly. The factor Ni takes into account the dose from immersion in the cloud of all the releases. The factor X/Q takes into account the dispersion of releases to the most conservative location. The factors are discussed in greater detail later. Beta air dose due to noble gases: ZiNi[(X/Q)v Qiv + (X/Q)s Qis] Where the constants have all been previously defined . Example Calculation Assume an analysis of the Stack and Vent Effluents indicate that 1.42E11 and 9.91E10 uCi of Xe-133 are released from each point respectively over the last month. This correlates to 1.81E4 and 1.26E4 4i/sec respectively. From Table 3-3, Ni is 1.05E3 mrad/yr per >>Ci/m3. (X/Q) for the Stack and Vent is 4.5E-8 and 2.0E-6 sec/m3 respectively. These values yield a beta air dose of 0.9 and 26.5 mrad/yr for the Stack and Vent respectively for a total of 27.4 mrad/yr or 6.8 mrad over the last quarter ~ The beta air dose limit due to noble gases is specified by TS Section 3.11.2.2. Organ Dose Due'o I-131, I-133, Tritium and Particulates with half-lives greater than 8 days. This calculation covers TS Section 3.11 .2.3, 3.11 .2.5, 4.11 .2.3, and
4. 11.2.5. 1. Organ dose due to I-131, I-133, Tritium and Particulates with half-lives greater than 8 days released is calculated monthly .
The factor Ri takes into account the dose received from the ground plane, inhalation, food (cow milk, cow meat and vegetation) pathways. Ws and Wv take into account the atmospheric dispersion from the release point to the location of the most conservative receptor for each of the respective pathways. The release is totaled over the period of concern. The factors are discussed in greater detail later. Organ dose due to iodine-131, iodine-133, tritium radionuclides in particulate form with half-lives greater than 8 days mrem ~ 3.17E-8 Ep [ Ei Rip [Ws Qis + Wv Qiv] ] Where:
3. 17E-8 Is the inverse of the number of seconds in a year Rip Is the factor that takes into account the dose to an individual organ from nuclide i through pathway p.
<<43- May 1986
Is the summation over all nuclides i. Is the summation over all pathways p, Ws, Wv Are the dispersion parameters for the stack and vent respectively for each pathway as appropriate sec/m or 1/m . See Table 3-22, Qis, Qiv Are the amount of activity of nuclide the stack or vent respectively over the period of i released from concern, uCi. If activity released is given in terms of release rate, uCi/sec, then the constant
3. 17E-8 sec may be omitted, provided that the annual dose calculated is divided by 4 to yield quarter dose, or 12. to yield monthly dose, as applicable.
Example Calculation Assume an analysis of the Stack and Vent Effluents indicate that 1.45E6 and 9.9E5 uCi of I-131 are released from each point respectively over the last quarter. This correlates 1.84E-1 and 1.26E-1 uCi/sec respectively. Calculate the dose to a childs organs. From Tables 3-8,11,13,16 and 19 the following table can be made: ORGAN Ri-GROUND Ri-I%M.ATION Ri-MILK Ri-MEAT Ri-VEGETATION or m2-mrem/ r mrem/ r m2-mrem/ r FACTOR uCi m3 uCi sec T BODY E .16E SKIN 2.09E7 BONE 4.81Z4 6.51E8 8.26E6 1.43E8 LIVER 4.81E4 6.55E8 8.32E6 1.44E8 THYROID 1.62E7 2. 17Ell 2.75E9 4.75 E10 KIDNEY 7. 88E4 1.08E9 1.37E7 2.36E8 LUNG G I-LLI 2. 84E3 5.83E7 7.40E5 1.28E7 Ws 1.34E-9 8.48E-9 3.64E-10 1. 15E-9 9.42E-10 Wv 2.90E-9 1.42E-7 4.73E-10 1.86E-9 1.50E-9 WsQs+WvQv 6. 12E-10 1 .95E-8 1.29E-10 4.46E-10 3.62E-10 From these values the following table of annual dose (mrem) can be calculated: ORGAN GROUND INHALATION MILK MEAT VEGE. TOTAL T BODY 1.11E-2 E-2 ~1E-3 MHE-2 K7KE-2 SKIN 1. 28E>>2 1.28E" 2 BONE 9.38E-4 8. 40E-2 3.69E-3 5. 18E-2 1.40E-1 LIVER 9.38E-4 8.45E-2 3.71E-3 5.21E-2 1.41E-1 THYROID 3. 16E-1 28. 0 1.23 17.2 46.7 KIDNEY 1.54E-3 1 .39E-1 6.11E-3 8.54E-2 2.32E-1 LUNG GI-LLI 5.54E-5 7.52E-3 3.30E-4 4.63E-3 1.25E-2 May 1986
In this case the maximum quarterly dose to the child organ is 46.7/4 11.7 mrem to the thyroid from I-131. The calculation would be repeated for all nuclides and age groups and summed to find the maximum dose to any organ. The dose limit to the maximum exposed organ is specified by TS Section 3.11.2.3 and 3.11.2.5. 3.4 Gaseous Effluent Dose Factor Definition and Derivation Bi andVi- Plume Shine Factor For Gamma and Beta Doses (Table 3-2) Bi (mrad/yr per uCi/sec) is calculated by modeling the effluent from the Stack as a line source with an elevation above ground equal to the stack height (131m). From "Introduction to Nuclear Engineering" by Lamarsh, page 410, the flux o at a point a distance of x from an .infinite line emitting S gammas/sec per cm is. o S/4x. S is proportional to release rate Q (uCi/sec) and inversly to wind speed U (cm/sec): S ~ Q/U. The distance of an individual on the ground from the elevated plume is approximately, equal to the height of the stack h (meters). The gamma radiation from the plume is attenuated by the air. This is proportional to the exponential of the negative product of the stack height h (m) and the air attenuation coefficient Uo, 1/m: exp (-Uo*h),. This is a conservative assumption .because only the portion of the plume directly overhead is at a distance of h. The bulk is much further away. Additionally, there is a dose buildup factor which, from RG 1.109 Appendix F-ll, 12, is equal to: 1+[(Uo-Ua)*Uo*h]/Ua where Ua (1/m) is the air energy absorption coefficient .
"4>" May 1986
The dose D at a point is proportional to the flux o, energy E (Mev) of the radiation, air energy absorption coefficient Ua (m-1) and unit conversion constant K: D K*o*E*Ua. Substitution in the above formula for flux from an infinite line source yields: D - K*S~E*Ua/[4*x]. Substitution for S yields: D K"Q*E~Ua/[4*x*U]. Substitution for x of Stack height h yields: D K*Q*E*Ua/[4*h*U]. Factoring in the air attenuation and corresponding dose buildup factors yields. D ~ K~Q*E"[Ua+(Uo-Ua)*Uo*h]exp(-Uo*h)/[4*h"U]. Bi is the gamma air dose received on the ground for a given release rate Q. Thus: B ~ D/Q ~ K*E*[Ua+(Uo-Ua)*Uo*h]~exp(-Uo~h)/[4*h*U]. Where. K ~ 1.447E4 mrad-dism /Mev~Ci-yr, U is 5.71 m/sec and the other symbols are as discussed above. To calculate Vi (mrem/yr per uCi/sec), the factor to account for the Total Body dose rate for a given release rate Q (uCi/sec) a conversion ratio of 1.1 mrem/mrad is assumed between tissue and air doses. If the Total Body tissue density Td (gm/cc) is assumed to be 5gm/cc (like a rock) and Ut (cm2/gm) is the energy absorption for tissue then: V 1.1*B*exp(-Td*Ut). Example Calculation Ua, Ue and Ut all vary with the energy of the radiation. Figure 3.5-6 and Table 3.5-1 (b-muscle) of the "CRC Handbook of Radiation Measurement and Protection" list values for the variables. For a 0.25 Mev gamma: Uo 0.0145 m-1 Ua ~ 0.0036 m-1 Ut ~ 0.0306 cm2/gm. May 1986
These values will yield a factor of 4.38E-3 and 4.14E-3 mrad, mrem/yr per uCi/sec respectively for B and V. Similarily for the primary energies of Xe135 the following table is obtainable: ENERGY YIELD B V MEV mrad/yr/uCi/sec mrem/yr/uCi/sec 0.25 0.9 4.38E-3 4.14E-3 0.6 0.03 9. 38E-3 8.77E-3 0.7 0.01 1.06E-2 9.97E-3 TOTALS FACTORING IN THE YEILDS: .31E- .07E-These values correspond to those listed on Table 3-2. It should be noted that only a limited number of nuclides are listed on Table 3-2. These are the most common noble gas nuclides encountered in effluents. If a nuclide is detected for which a factor is not= listed, then ODCM. it will be calculated and included in a revision to the Semi-Infinite Cloud Immersion Dose Factors (Table 3-3) Ki, Li, Mi and Ni are the factors which take into account the dose from immersion in the semi-infinite cloud of gaseous releases. These are taken from RG 1.109, Table B-l, and multiplied by lE6 to convert, from units of mrem,mrad/yr per pCi/m3 to mrem,mrad/yr per uCi/m3. Dose Rate Factor for I-131, I"133, Tritium 'and Particulates with Half"lives greater than 8 days. Table 3-4 Ground Plane Pi (m2~rem/yr per uCi/sec) takes into account several factors among these are the dose rate to the total body from exposure to radiation deposited on the ground. (From NUREG 0133, section 5.2.1.2) INSERT SYMBOLS Where constant of unit coversion, 106 pCi/pCi. K" a constant of unit conversion, 8760 hr/year.
>i ~ the decay constant for the ith radionculide, sec t ~ the exposure period, 3.15 x 107 sec (1 year) .
DFQ the ground plane dose conversion factor the the ith radionuclide (mrem/hr per pCi/m ) . The deposition rate onto the ground plane results in a ground plane concentration that is assumed to persist over a year with radiological decay the only operating removal mechanism for each radionuclide. . The ground plane dose conversion factors for the ith radionuoldde, Dpdi, are presented in 1able E-6 of Regulatory Guide 1.109, in units of mrem/hr per pCi/m May 1986
Resolution of the units yields. Pi (Ground) 8.76 x 10 DFQ (1-e "i )/ki Example Calculation For the I-131 total body dose rate factor for exposure from the ground: 9.98E-7 sec-1 DFGi 2.80E-9 mrem/hr per Ci/m2 These values will yield a Pi factor of 2.46E7 m2~rem/yr per uCi/sec as listed on Table 3-4. It should be noted that only a limited number of nuclides are listed on Table 3-4. These are the mos t common nuclides encountered in effluents. If a nuclide is detected for which a factor 'is not listed, then included in a revision to the ODCM. it will be calculated and Pi (m2~rem/yr per uCi/sec) also takes into account the dose rate to the skin from exposure to the ground. Example Calculation For the I-131 skin dose rate factor for exposure from the ground:
~ 9.98E-7 sec"1 DFGi 3.40E-9 mrem/hr per pCi/m2 These values will yield a Pi factor of 2.98E7 m2-mrem/yr per uCi/sec as listed on Table 3-4. It should be noted that only a limited nubmer of nuclides are listed on Table 3-4. These are the most common nuclides encountered in effluents. If a nuclide is. detected for which a factor is not listed, then ith will be calculated and included in a revision to the ODCM.
Table 3-5, Inhalation Pi (mrem/yr per uCi/m3) also takes into account the dose rate to various organs from inhalation exposures (From NUREG 0133, section 5.2.1. 1) Pi ~ K'(BR) DFAi (mrem/yr per @CD./m ) Where. K' constant of unit conversion, 106 pCi/XCi. BR the breathing rate of the infant age group, in m3/yr. DFAi ~ the organ inhalation dose factor for the infant age group for the ith radionuclide, in mrem/pCi. The total body is considered as an-organ in the selection of DFAi. May 1986
The age group considered is the infant group. The infant's breathing r ate is taken as 1400 m3 /yr rom Table E-5 of Regulatory Guidef 1.109. The inhalation dose factors for the infant, DFAi are presented in Table E-10 o Regulatory Guide f l. 109, in unit s o f mrem/pCi .. Resolution of the units yeilds.'i (inhalation) 1.4 x 10 DFAi. Example Calculation
.'or the I-131 thyroid dose rate factor for exposure from inhalation:
DFAi 1.06E-2 mrem per pCi This value will yield a Pi factor of 1.48E7 mrem/yr per uCi/m3 as listed on Table 3-5. It should be noted that only a limited number of nuclides are listed on Table 3-5. These are the most common nuclides encountered in effluents. which a factor is not listed, then it will be calculated and included If a nuclide is detected f'r in a revision to the ODCM. Table 3-6, Food (Cow Milk) Pi (m2~rem/yr per uCi/sec) also takes into account the dose rate to various organs from the ingestion of cow milk. (From NUREG 5.2.1.3). 0133'ection O(U )
~ ~
K r > +z F DFL1 ~t
-A 3
t f] (a> mree/yr per uC1/sec) p ( w Where: K' constant of unit conversion, 106 pCi/pCi.
~ the cow's consumption rate, in kg/day (wet weight).
Uap the infant ' milk consumption rate, in 1 iters/yr. Yp ~ the agricultural productivity by unit area, in kg/m
~ the stable element transfer coefficients, Fm in days/liter.
r fraction of deposited activity retained on cow's feed grass. DFLi the maximum organ ingestion radionuclide, in mrem/pCi . dose factor for the ith 1i ~ the decay constant for the ith radionuclide, in sec the decay constant for removal of activity on leaf and plant surfaces by weathering, 5.73 x 10 7 sec
.( corresponding to a 14 day half-time).
the transport time from pasture to cow, to milk, to infant, in sec. May 1986
A fraction of the airborne deposition is captured by the ground plane vegetation cover. The captured material is removed from the vegetation (grass) by both radiological decay and weathering processes. The values of Qp, U n, and Y are provided in Regulatory Guide 1 .109, Tables E-3, E-S, and E-E5, as 50 kg/day, 330 liters/day and 0.7 kg/m , respectively. The value Guide 1.109, Table E-15, as 2 days (1.73 x 105 seconds) tf is provided in Regulatory
~ The fraction, r, has a value of 1.0 for radioiodines and 0.2 for particulates, as presented in Regulatory Guide 1.109, Table E-15.
Table E-1 of Regulatory Guide 1.109 provides the stable element transfer coefficients, Fm, and Table E-14 provides the ingestion dose factors, DFLi, for the infant's organs . Resolution of the units yields:
~<:, tooer z.4xl(P o 0R.< [e 1 f] (m'~rem/yr per ~C1/sec) for all radionuclides, except tritium.
The concentration of tritium in milk is based on its airborne concentration rather than the deposition rate. K'" QFU OFl.< f0.7 (0. / )) (mree/yr per pCi/ms) Where: Ka constant of unit conversion, 10 gm/kg ~ H absolute humidity of the atmosphere, in gm/m / 0.75 ~ the fraction of total feed that is water. 0.5 the ration of the specific activity of the feed grass water to atmospheric water. From Table E-1 and E-14 of Regulatory Guide 1.109, the values of F and DFLi for tritium are 1 .0 x 10 day/liter and 3.08 x 10 ~ mrem per pCi, respectively. Assuming an average absolute humidity of 8 grams/meter , the resolution of units yields: Pi (food) ~ 2.4 x 103 mrem/yr per pCi/m for tritium, only Example Calculation: For I-131 thyroid does rate factor for exposure from cow milk ingestion:
- May 1986
r 1.0 unities s for Iodine s Fm ~ 6E-3 days/liter DFLi 1.39E-2 mre m/pCi 9.98E-7 sec-1
~w 5.73E-7 sec-1 tf 1.73E+5 sec These values will yield a Pi factor of 1.07E12 mrem/yr per uCi/sec as listed on Table 3-6. It should be noted that only a limited number of nuclides are listed on Table 3-6. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then in a revision to the ODCM.
it will be calculated and included 3.4.4 Dose Factor for I-131, I-133, Tritium and Particulates with half-lives greater than 8 days. TABLES 3.7 to 3.10, Ri VALUES - INHALATION Ri (mrem/yr per uCi/m3) takes into account several factors, among these are the dose rate to various organs from inhalation exposure. (From NUREG 0133, Section 5.3.1.1). Ri K'(BR) a (DFAi)a- (mrem/yr per uCi/m ) Where: K' constant of unit conversion, 10 pCi/pCi ~ (BR)a th~ breathing rate of the receptor of age group (a),in m /yr. (DFQ)a ~ the organ inhalation dose factor for the receptor of age group (a) for the ith radionuclide, in mrem/pCi. The total body is considered as an organ in the selectS, on of (DFAi) a. The breathing rates (BR)a for the various age groups are tablulated below, as given in Table E-5 of the Regulatory Guide 1.109 Rate (m 'reathin
~/ r)
Infant 140 0 Child 3700 Teen 8000 Adult 8000 Inhalation dose factors (DFAi)a for the various age groups are given in Tables E-7 throught E-10 of Regulatory Guide 1.109. Example Calculation: For the I-131 infant thyroid dose factor for exposure from inhalation: DFAi ~ 1.06E-2 mrem per pCi May 1986
These values will yield a Ri factor of 1.48E7 mrem/yr per uCi/m3 as listed on Table 3-7. It should be noted that only a limited number of nuclides are listed on Table 3-7 thru 3-10. These are the most common nuclides encountered in effluents If a nuclide is detected it
~
for which a factor is not listed, then will be calculated nd included in a revision to the ODCM. TABLE 3-11, Ri VALUES GROUND PLANE Ri (m2~rem/yr per uCi/sec) also takes into account the dose from exposure to radiation deposited on the ground. (From NUREG 0133, Section 5.3. 1.2) ~ K'K"(SF)DFQ [(1-e"i")/~i) (m mrem/yr per uCi/sec) Where: K' constant of unit conversion, 10 pCi/pCi. K" ~ a constant of unit conversion, 8760 hr/year.
~i ~ the decay constant for the ith radionuclide, sec 1 .
t the exposure time, 4.73 x 10 sec (15 years) ~ DFGi the ground plane dose conversion factor for the ith radionuclide (mrem/hr per pCi/m ) . SF ~ the shielding factor (dimensionless). A shielding factor of 0.7 is suggested in Table E-15 of Regulatory Guide 1.109. A tabulation of DFGi values is presented in Table E-6 of Regulatory Guide 1.109. Example Calculation: For the I-131 total body dose factor for exposure to the ground: 9.98E-7 sec-1 DFGi ~ 2.80E-9 mrem/hr per pCi/m2 These values will yield a Ri factor of 1.72E7 m2mrem/yr per uCi/sec a s listed on Table 3 11. It should be noted that only a limited number of nuclides are listed on Table 3-11. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then it will be calculated and included in a revision of the ODCM. Ri (m2~rem/yr per uCi/sec) also takes into account the dose to the skin from exposure to the ground. May 1986
Example Calculation: For the I-131 skin dose factor for exposure to the ground: 9.98E-7 sec-1 DFGi 3.40E-9 mrem/hr per pCi/m2. These values will yield a Ri factor of 2.09E7 m2~rem/yr per uCi/sec as listed on Table 3 11. It should be noted that only a limited number of nuclides are listed on Table 3-11. These are the most itIfwill common nuclides encountered in effluents. a nuclide is detected for which a factor is not listed, then be calculated an included i'n a revision to the ODCM. TABLES 3-12 to 3-15 Ri VALUES COW MILK Ri (m2mrem/yr per uCi/sec) also takes into account the dose rate to various organs from the ingestion of milk for all age groups. (From NUREG 0133, Section 5.3.1.3) ~ 4(Ua ~ r~r (>-r r >e-'i'~
]
t 5 2 (0 ~~/W ptr uC)/etc} Where: K' a constant of unit conversion, 10 pCi/pCi ~ +- the cow's consumption rate, in kg/day (wet weight). Uap the receptor' mi lk consumption rate, in 1 iters/yr. Y P the agricultural productivity by unit area of pasture feed grass, in kg/m2 Ys the agricultural productivity by unit area of stored feed, in kg/m2. Fm
~ the stable element transfer coefficients, in days/liter.
r ~ fraction of deposited activity retained on cow's feed grass. (DFLi)a the organ ingestion dose factor for the ith radionuclide for the receptor in age group (a), in mrem/pCi ~ the decay constant for the ith radionuclide, in sec May 1986
"w the decay constant for removal of activity on leaf and plant surfaces by weathering, 5.73 x 10 " sec (corresponding to a 14 day half-time) ~ . the transport time from pasture to cow, to milk, to 'eceptor, in sec.
th the transport time from pasture, to harvest, to cow, to milk, to receptor, in sec. fP mm fraction of the year that the cow is on pasture (dimensionless). fs me fraction of the cow feed that is pasture grass while the cow is on pasture (dimensionless). SPECIAL NOTE: The above equation is applicable in the case that the milk animal is a goat. Milk cattle are considered to be fed 'from two potential sources, pasture grass and stored feeds. Following the development in Regulatory Guide 1.109, the values of fp and fs will be considered unity. Tabulated below are the appropriate parameter values and their reference,to Regulatory Guide 1.109. In case that the milk animal is a goat, rather than a cow, refer to Regulatory Guide 1.109 for the appropriate parameter values. Parameter Value Fable r (dimensionless) 1.0 for radioiodine E-15 0.2 for particulates E-15 Fm (days/liter) Each stable element E-1 Uap (liters/yr) - Infant 330 E-5 Child 330 E-5
- Teen 400 E-5 Adult . 310 E-5 (DFLL) a fmzem/PCi) Each radionuclide E-ll to E-14 Yp (kg/m ) 0.7 E-15 Ys (kg/m2) 2.0 E-15 tf (seconds) 1.73 x 105 (2 days) E-15 th (seconds) 7.78 x 10 (90 days) E-15 QF (kg/day) 50 E-3 The concentration of tritium in milk is based on the .airborne concentration rather than the deposition. Therefore, the Ri is based on [x/Q]:
FgFUapFLi[0.75(0.5/H)) (mrem/yr per uCi/m ) May 1986
Where: K" a constant of unit conversion, 10 gm/kg ~ H absolute humidity of the atmosphere, in gm/m3 0.75 ~ the fraction of total feed that is water. 0.5 ~ the ratio of the specific activity of the feed grass water to atmospheric water. and other parameters and values are given above. The value of H is considered as 8 grams/meter , in lieu of site specific information. Example Calculation: For I-131 infant thyroid dose factor from milk ingestion: r ~ 1 .0 unitless for Iodines Fm 6 E-3 days/liter for cows and 6E-2 for goats DFLi 1.39E-2 mrem/pCi
~ 9.98E-7 sec -1 ~w 5.73E-7 sec -1 tf 1.73E+5 sec.
These values will yield a factor of 5.26E11 and 6.31Ell mrem/yr per uCi/sec respectively for cow and goat milk. However, the actual dose to the infant thyroid is also dependant on the highest relative deposition at respective cow and goat locations. At the Nine Mile Point Nuclear Station these deposition coefficients are 4.73E-10 and 1.33E-10 m-2 respectively for cows and goats'ecause the goat deposition is relatively so much smaller than the slightly larger Ri factor, cow milk is the limiting milk If the location of the cow
~
and goat milk receptors changes so that this is no longer true then the Ri factor will be revised accordingly. Table 3 12 list the infant thyroid dose factor from I-131 as 5.26E11 mrem/yr per uCi/sec ~ It should be noted that only a limited number of nuclides are listed on Table 3 12 thru 3 15. These are the most common . nuclides encountered in effluents If a nuclide is detected for which a factor is not listed, then in a revision to the ODCM.
~
it will be calculated and included TABLES 3-16 3-18, Ri VALUES COW MEAT (m2~rem/yr per uCi/sec) also takes into account the dose rate to various organs from the ingestion of cowmeat for all age groups except infant. (From NUREG 0133, Section 5.3.1 .4) R<[D/Ql K' (m 2 q~(u, ) F (r)(WL<) 1 a wren/~", oar uc1/sec)
~ (1 ff )e S
3 h e
-X<t<
May 1986
Where: Ff the stable element transfer coefficients, in days/kg. Uap the receptor' meat consumption rate for age (a), in kg/yr. tf the transport time from pasture to receptor, in sec. th the transport time from crop field to receptor, in sec. Tabulated below are the appropriate parameter values and their reference to Regulatory Guide 1.109. Parameter Value Table(RG1.109) r (dimensionless) 1.0 for radioiodine E-15 0.2 for particulates E-15 Ff (days/kg Each stable element E-1 Uap (kg/yr Infant 0 E-5
- Child 41 E-5 - Teen 65 E-5 Adult 110 E-5 (DFLi) a (.mrem/pCi) Each radionuclide E-11 to E-14 Y (kg/mc) 0.7 E-15 Ys (kg/m ) 2.0 E-15 tf (seconds) 1.73 x 10 (20 days) E-15 th (seconds) 7.78 x 10 (90 days) E-15 Qp (kg/day) 50 E-3 The concentration of tritium in meat is based on the airborne concentration rather than the deposition. Therefore, the Ri is based on [x/9]: <'<'"F~IFU (OFL) f0.75(0.5/H)j (mrna/yr Per l,gl/H) a where all terms are defined above in this manual ~
Example Calculation: For I-131 child thyroid dose factor from cow meat ingestion. Ff 2.9E-3 days r 1.0 unitless for Iodines DFLi ~ 5.72E-3 mrem/pCi ~ These values will yield a Ri factor of 2.75E9 m2~rem/yr per uCi/sec as listed on Table 3 16. It should be noted that only a limited number of nuclides are listed on Table 3-16 thru 3-18. These are the it If a nuclide is most common nuclides encountered in effluents. detected for which a factor is not listed, then will be calculated in a revision to the ODCM. May 1986
TABLES 3-19 to 3-21 Ri VALUES - VEGETATION Ri (m2~rem/yr per uCi/sec) also takes into account the dose to various organs from the ingestion of vegetation for all age groups except infant ~ (From NUREG 0133, Section 5.3.1.5) ~ The integrated concentration in vegetation consumed by man follows the expression developed in the derivation of the milk factor. Man is considered to consume two types of vegetation (fresh and stored) that differ only in the time period between harvest and consumption, therefore: l >- (0FL 'I ) ULf 1 L+ USf [ Ai+ X a aL e ag J (m 2 m rem/3 r per pCI /s ec ) K' constant of unit conversion, 10 pC1/PC1 ~ L Ua the consumption rate of fresh leafy vegetat1on by the receptor 1n age group (a), 1n kg/yr. U ~ the consumption rate of stored vegetation by the receptor in age group (a). 1n kg/yr. fL ~ the fraction of the annual intake of fresh leafy vegetation grown ilocally . the fraction of the annual intake of stored vegetation green locally. the average t1me behreen harvest of leafy vegetation and its consunptlon, 1n seconds. t the average t1me bebaen harvest of stored vegetation and in seconds. its consumption, Yv the vegetat1on areal density. in kg/ms. and all other factors are defined in this manuals Tabulated below are the appropriate parameter values and their reference to Regulatory Guide 1.109. Parameter Value Table r (dimens ionless) 1 ~ 0 for radioiodines E-1 0.2 for particulates E-1 (DFLi ) a (mrem/pC1) Each radionucl1de E-11 to E-14 U (kg/yr) - Infant 0 E-5
- Child 26 E-5 - Teen 42 E-5 - Adult 64 E-5 U (kg/yr) - Infant 0 E-5 - Child 520 E-5 - Teen 630 E-5 - Adult 520 E-5 .fL (dimensionless) s1te specific (default ~ 1.0) f (dimensionless) site specific (default ~ '0.76) (see RGIJ08page 28) tL (seconds) 8.6 X 10 (1 day) E-15 th (seconds) 5.18 X 10 (60 days) E-15 Y (kg/m ) 2.0 E-15 May 1986
The concentration of tritium in vegetation is based on the airborne concentration rather than the deposition. Therefore, the Ri is based on [x/Q]: "(x/Ql . K'r' u r a L
+ u e (OFL< a f' .5/H)] (cree/yr per a < usaf/e ).
where all terms have been defined above and in this manual ~ Example Calculation For I-131 child thyroid dose factor to the from vegetation ingestion.'
~ 1 .0 unitless for Iodines DFLi ~ 5.72E-3 mrem.pCi.
These values will yield a Ri factors o f 4.75E10 m2mrem/yr pe r uCi/sec as listed on Table 3 19. It should be noted that only a limited number of nuclides are listed on Table 3-19 thru 3-21. These are the most common nuclides encountered in effluents. If a nuclide is detected for which a factor is not listed, then calculated and included in a revision to the ODCM. it will be X/Q and Wv Dispersion Parameters for Dose Rate, Table 3-22 The dispersion parameters for the whole body and skin dose rat e calculation correspond to the highest annual average dispersion parameters at or beyond the unrestricted area boundary. This is at the East Site boundary. These values were obtained from the Nine Mile Point Unit 2 Final Environmental Statement, NUREG 1085 Table D-2 for the Vent and stack. These were calculated using the methodology of Regulatory Guide 1.111, Rev. 1. The Stack was modeled as an elevated release point because its height is more than 2.5 times than any adjacent building. The Vent was modeled as a ground level release because even though it it is not more than 2.5 times the height. is higher than any adjacent building The NRC Final Environmental Statement values for the Site Boundary X/Q and D/Q terms were selected for use in calculating Effluent Monitor Alarm Points and compliance with Site Boundary Dose Rate specifications because they are conservative when compared with the corresponding NMPC Environmental Report values. In addition, the Stack "intermittent release" X/Q was selected in lieu of the "continuous" value, since it is slightly larger, and also would allow not making a distinction between long term and short term releases . The dispersion parameters for the organ dose calculations were obtained from the Environmental Report, Figures 7B-4 (Stack) and 7B-8 (Vent) by locating values corresponding to currently existing (1985) pathways. It should be noted that the most conservative pathways do not all exist at, the same location. It is conservative to assume that a single individual would actually be at each of the receptor locations. May 1986
3.4.6 Wv and Ws Dispersion Parameters for Dose, Table 3-22 The dispersion parameters for dose calculations were obtained chiefly from the Nine Mile Point Unit 2 Environmental Report Appendix 7B3 as noted in Section 3.4.5. These were calculated using the methodology of Regulatory Guide 1.111 and NUREG 0324. The Stack was modeled as an elevated release point because its height is more than 2.5 times than any adjacent building. The Vent was modeled as a combined elevated/ground level release because even though any adjacent building it it is higher than is not more than 2.5 'times the height . Average meterology over the appropriate time period was used. Dispersion parameters not available from the ER were obtained from C.T. Main Data report dated November, 1985, or as described in Section 3.4.5, the FES. 3.5 I-133 Es timation The 'Stack and Vent Effluent Monitor at Nine Mile Point-Unit 2 are on line isotopic monitors. They are designed to automatically collect iodine. samples on charcoal cartridges and isotopically analyze them with a sensitivity which exceeds the LLD requirement on TS Table
4. 11-2 of 1E-12 uCi/cc. During those time periods in which the I-133 analysis cannot meet the LLD requirement, the I-133 concentration will be estimated as 4 times the I-131 concentration, or by ratio applied to the I-1'31 concentration. The ratio will be determined at least quarterly by analysis of short duration samples .
3.6 Use of Concurrent Meteorological Data vs. Historical Data It is the intent of NMPC to use dispersion parameters based on historical meteorological data to set alarm points and to determine or predict dose and dose rates in the environment due to gaseous effluents. When the methodology becomes available, it is the intent to use meteorological conditions concurrent with the time of release to determine gaseous pathway doses. Alarm points and dose predictions or estimates will still be based on historical data. The ODCM will be revised at that time. 3.7 Gaseous Radwaste Treatment S stem 0 eration Technical Specification 3.11.2.4 requires the Gaseous Radwaste Treatment System to be in operation whenever the main condenser air effector system is in operation. Since the system was designed without a bypass, station design results in compliance with the specification. The components of the system which must operate to treat offgas are the Preheater, Recombiner, Condenser, Dryer, Charcoal Adsorbers, HEPA Filter, and Vacuum Pump. See Figures 3-1, 3-2, and 3-3, Offgas System. May 1986
3.8 Ventilation Exhaust Treatment S stem 0 eration Technical Specification 3.11.2.5 requires the Ventilation Exhaust Treatment System to be OPERABLE when projected doses in 31 days due to iodine and particulate releases would exceed 0.3 mrem to any organ of a member of the publica The appropriate components, which affect iodine or'articulate release, to be OPERABLE are:
1) HEPA Filter -Radwaste Decon Area

2) HEPA Filter Radwaste Equipment Area

3) HEPA Filter Radwaste General Area Whenever one of these filters is not OPERABLE, iodine and particulate dose projections will be made for the remainder of the current calendar month, and for each month (at the time of calculating cumulative monthly dose contributions) that the filter remains inoperable, in accordance with 4.11.2.5. 1. Predicted release rate will be used, with the methodology of Section 3.3.3. See Figure 3-5, Gaseous Radiation Monitoring.
May 1986
TABLE 3-1 OFFGAS PRETREATMENT* DETECTOR RESPONS E NUCLIDE NET CPM/pCi/cc Kr 85 4.30E+3 Kr 85m 4.80E+3 Kr 87 8.00E+3 Kr 88 7.60E+3 Xe 133 1.75E+3 Xe 133m Xe 135 5.10E+3 Xe 135m Xe 137 8.10E+3 Xe 138 7. 10E+3 <<Values from SWEC purchase specification NMP2-P281F May 1986
TABLE 3-2 PLUME SHINE PARAMETERS* NUCLIDE B (mrad/ r
. uCi/sec) V (mrem/yr . pCi/sec)
K" 83m 3.5. 1E-5 3.28E-5 Kr 85 3. 39E-3 3.21E-3 Kr 85m 1.04E-2 9.98E-3 Kr 87 2.34E-2 2.21E-2 Kr 88 2.01E-2 1.92E-2 Kr 89 1.59E-2 1. 51E-2 Xe 131m 6.90E-5 6.55E-5 Xe 133 6. 12E-4 5.93E-4 Xe 133m 3.62E-4 3.44E-4 Xe 135 4.31E-3 4.09E-3 Xe 135m 6.55E-3 6.12E-3 Xe 137 3.07E-3 2.88E-3 Xe 138 1.38E-2 1.33E-2 Ar 41 1.69E-2 1.61E-2
*Bi and Vi are calculated for critical site boundary location; 1 .6km in the easterly direction.
May 1986
TABLE 3-3 DOSE FACTORS*
~N(B-Air Nuclide ~K(r Bcdy)** Li (B-Skin)** ~M(y-Air)*** ) ***
Kr 83m 7.56E-02 1.93E1 2.88E2 Kr 85m 1.17E3 1.46E3 1.23E3 1.97E3 Kr 85 '1.61E1 1.34E3 1.72El 1.95E3 Kr 87 5. 92E3 9.73E3 6. 17E3 1.03E4 Kr 88 1.47E4 2.37E3 1.52E4 2.93E3 Kr 89 1.66E4 1 ~ 01E4 1.73E4 1.06E4 Kr 90 1.56E4 7.29E3 1.63E4 7.83E3 Xe 131m 9. 15E1 4.76E2 1.56E2 1.11E3 Xe 133m 2.51E2 9.94E2 3.27E2 1.48E3 Xe 133 2.94E2 3.06E2 3. 53E2 1.05E3 Xe 135m 3. 12E3 7.11E2 3.36E3 7.39E2 Xe 135 l. 81E3 1.86E3 1.92E3 2.46E3 Xe 137 1.42E3 1.22E4 1.51E3 1.27E4 Xe 138 8.83E3 4.13E3 9.21E3 4. 75E3 Ar 41 8.84E3 2.69E3 9.30E3 3.28E3 "From, Table B-l.Regulatory Guide 1.109 Rev. 1
**mrem/yr per uCi/m 3 .
mrad/yr per uCi/m 3 May 1986
TABLE 3-4 Pi VALUES GR UND PLANE**
!iCi/se c NUCLIDE TOTAL BODY SKIN H 3 C 14 Cr 51 6.64E6 7.85E6 Mn 54 1.10E9 1.29E9 Fe 59 3.88E8 4.56E8 Co 58 5.27E8 6. 18E8 Co 60 4.40E9 5. 17E9 Zn 65 6.87E8 7.90E8 Sr 89 3.06E4 3.56E4 Sr 90 Zr 95 3.44E8 3.99E8
Nb 95 3.50E8 4. 12E8 Mo 99 5.71E6 6.61E6 I 131 2.46E7 2.98E7 I 133 3.50E6 4.26E6 Cs 134 2. 81E9 3.28E9 Cs 137 1.15E9 1.34E9 Ba 140 2. 93E7 3.35E7

La 140 2. 10E8 2.38E8 Ce 141 1.95E7 2. 20E7 Ce 144 5.85E7 6.77E7

Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.

~Calculated in accordance with NUREG 0133, Section 5.2.1.2.
May 1986
TABLE 3-5 P VALUES INHALATION+* m~rem/ r 3 uci/m NUCLIDE BONE LIVER Y. BODY YHYROID KIDNEY LUNG GI-LLI H 3 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 C 14 2.65E4 5.31E3 5.31E3 5.31E3 5.31E3 5.31E3 5.31E3 Cr 51 8.95E1 5.75E1 1.32E1 1.28E4 3.57E2 Mn 54 2.53E4 4.98E3 4.98E3 1.00E6 7.06E3 Fe 59 1.36E4 2.35E4 9.48E3 1.02E6 2.48E4 Co 58 1.22E3 1.82E3 7. 77E5 1'lE4 Co 60 8.02E3 1.18E4 4.51E6 3. 19E4 Zn 65 1.93E4 6.26E4 3.11E4 3.25E4 6.47E5 5.14E4 Sr 89 3.98E5 1.14E4 2.03E6 6.40E4 Sr 90 4.09E7 2.59E6 1.12E7 1-31E5 Zr 95 1.15E5 2.79E4 2.03E4 3.11E4 1.75E6 2.17E4
Nb 95 1.57E4 6.43E3 3.78E3 4.72E3 4.79E5 1.27E4 Mo 99 1.65E2 3.23El 2.65E2 1.35E5 4.87E4 I 131 3.79E4 4.44E4 1.96E4 1.48E7 5.18E4 1.06E3 I 133 1.32E4 1.92E4 5.60E3 3.56E6 2.24E4 2. 16E3 Cs 134 3.96E5 7.03E5 7.45E4 1.90E5 7.97E4 1.33E3 Cs 137 5.49E5 6.12E5 4.55E4 1.72E5 7.13E4 1.33E3 Ba 140 5.60E4 5.60El 2. 90E3 1.34E1 1.60E6 3.84E4

La 140 5.05E2 2.00E2 5.15El 1.68E5 8.48E4 Ce 141 2.77E4 1.67E4 1.99E3 5.25E3 5.17E5 2.16E4 Ce 144 3.19E6 1.21E6 1.76E5 5.38E5 9.84E6 1.48E5

Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
""Calculated in accordance with NUREG 0133, Section 5.2. 1.1. May 1986
TABLE 3-6 P i VALUES FOOD (Cow 2 Milk)*+* m mrem/yr uCi/se c NU GLIDE BONE LIVER Y ~ BODY YHYEOYD KIDNEY LUNG GI-LLI
*H 3 2.40E3 2.40E3 2.40E3 2.40E3 2.40E3 2.40E3 *C 14 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 1.64E5 1.07E5 2.34E4 2.08E5 4.78E6 Mn 54 3.97E7 8. 99E6 8.80E6 1.46E7 Fe 59 2.28E8 3.99ES 1.57ES 1.18ES 1.91ES Co 58 2.47E7 6.16E7 6. 15E7 Co 60 8.98E7 2.12ES 2.14E8 Zn 65 5.65E9 1.94E10 8.94E9 9.40E9 l. 64E10 Sr 89 1.28E10 3.67E8 2.63ES Sr 90 1.24E11 3. 15E10 1.55E9 Zr 95 6.93E3 1 .69E3 1.20E3 1.82E3 8.41E5
"Nb 95 7.07E5 2.91E5 1.68E5 2.09E5 2.46ES Mo 99 2.12E8 4.13E7 3. 17E8 6.98E7 I 131 2.77E9 3.26E9 1.43E9 1.07E12 3.81E9 1.16E8 I 133 3.69E7 5.37E7 1.57E7 9.77E9 6.31E7 9.09E6 Cs 134 3.71E10 6.92E10 6.99E9 1.78E10 7.31E9 1.88E8 Cs 137 5.24E10 6.13E10 4.35E9 1.65E10 6.67E9 1.92ES Ba 140 2.45E8 2.45E5 1.26E7 5.83E4 1.51E5 6.03E7

<<La 140 3.79E2 1.49E2 3.84El 1.75E6 Ce 141 4.41E4 2.69E4 3.17E3 8.30E3 1.39E7 Ce 144 2.37E6 9.69E5 1N33E5 3.92E5 1.36ES
*mrem/yr per uCi/m3. ~*Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
Calculated in accordance with NUREG 0133 Section 5.2.1.3.
May 1586
~ ~ ~ ~
TABLE 3-7 R i VALUES . ZÃiALATION INFANT** m~rem/ r 3 uci/m NU CLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG G I-LLI H 3 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 6.47E2 C. 14 2. 65E4 5. 31E3 5.31E3 5.31E3 5.31E3 5. 31E3 5. 31E3 Cr 51 8.95El 5.75E1 1.32E1 1.28E4 3.57E2 Mn 54 2.53E4 4.98E3 4.98E3 1.00E6 7.06E3 Fe 59 1.36E4 2.35E4 9.48E3 1.02E6 2.48E4 Co 58 1.22E3 1.82E3 7. 77E5 1.11E4 Co 60 8.02E3 1.18E4 4.51E6 3. 19E4 Zn 65 1.93E4 6.26E4 3.11E4 3.25E4 6.47E5 5.14E4 Sr 89 3.98E5 1.14E4 2.03E6 6.40E4 Sr 90 4.09E7 -2.59E6 1.12E7 1.31E5 Zr 95 1.15E5 2.79E4 2.03E4 3. 11E4 1.75E6 2. 17E4 "Nb 95 1.57E4 6.43E3 3.78E3 4.72E3 4.79E5 1.27E4 Mo 99 1.65E2 3.23E1 2.65E2 1.35E5 4.87E4 I-131 3.79E4 4.44E4 1.96E4 1.48E7 5. 18E4, 1.06E3 I 133 1.32E4 1.92E4 5.60E3 3.56E6 2.24E4 2. 16E3 Cs 134 3.96E5 7.03E5 7.45E4 1.90E5 7.97E4 1.33E3 Cs 137 5.49E5 6. 12E5 4.55E4 1 .72E5 7.13E4 1 e33E3 Ba 140 5.60E4 5.60E1 2.90E3 1.34E1 1.60E6 3.84E4 "La 140 5.05E2 2.00E2 5.15E1 1.68E5 8.48E4 Ce 141 2.77E4 1.67E4 1.99E3 5.25E3 5.17E5 2.16E4 Ce 144 3.19E6 1.21E6 1.76E5 5.38E5 9.84E6 1.48E5
Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
<<*This and following R Tables Calculated in accordance with NUREG 0133, Section 5.3. 1, excpet C 14 values in accordance with Regulatory Guide 1.109 Equation C-8. May 1986
TABLE 3-8 R VALUES INHALATION - CHILD m~reml r 3 nCi/m NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI H 3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 1.12E3 C 14 3.59E4 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 6.73E3 Cr 51 1.54E2 8.55El 2.43E1 1.70E4 1 .08E3 Mn 54 4.29E4 9.51E3 1.00E4 1-58E6 2.29E4 Pe 59 2.07E4 3.34E4 1.67E4 1.27E6 7.07E4 Co 58 1.77E3 3.16E3 1. 11E6 3. 44E4 Co 60 1.31E4 2.26E4 7.07E6 9.62E4 Zn 65 4.26E4 1.13E5 7.03E4 7.14E4 9.95E5 1.63E4 Sr 89 5.99E5 1.72E4 2. 16E6 1 .67E5 Sr 90 1.01E8 6.44E6 1.48E7 3.43E5 Zr 95 1.90E5 4.18E4 3.70E4 5.96E4 2.23E6 6-11E4
Nb 95 2.35E4 9. 18E3 6.55E3 8.62E3 6.14E5 3.70E4 Mo 99 1.72E2 4.26El 3.92E2 1.35E5 1.27E5 I 131 4.81E4 4.81E4 2.73E4 1.62E7 7.88E4 2.84E3 I 133 1.66E4 2.03E4 7.70E3 3.85E6 3.38E4 5.48E3 Cs 134 6. 51E5 1.01E6 2.25E5 3.30E5 1.21E5 3. 85E3 Cs 137 9.07E5 8.25E5 1.28E5 2.82E5 1 .04E5 3.62E3 Ba 140 7.40E4 6.48E1 4.33E3 2. 11El 1.74E6 1.02E5

La 140 6.44E2 2. 25E2 7.55El 1.83E5 2.26E5 Ce 141 3.92E4 1.95E4 2.90E3 8. 55E3 5.44E5 5. 66E4 Ce 144 6.77E6 2. 12E6 3.61E5 1.17E6 1.20E7 3.89E5

Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-9 R VALUES INHALATION TEEN m~rem/ r 3 uCi/m NUCLIDE BONE LIVER I. BODY IHYEOID KIDNEY LUNG GI-LLI H 3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 '1.27E3 C 14 2.60E4 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 Cr 51 1.35E2 7.50E1 3.07E1 2.10E4 3.00E3 Mn 54 5.11E4 8.40E3 1.27E4 1.98E6 6.68E4 Fe 59 1.59E4 3.70E4 1.43E4 1.53E6 1.78E5 Co 58 2.07E3 2.78E3 1.34E6 9.52E4 Co 60 1.51E4 1.98E4 8.72E6 2.59E5 Zn 65 3.86E4 1.34E5 6.24E4 8.64E4 1.24E6 4.66E4 Sr 89 4.34E5 1.25E4 2.42E6 3.71E5 Sr 90 1.08E8 6.68E6 1.65E7 7.65E5 Zr 95 1.46E5 4.58E4 3.15E4 6.74E4 2.69E6 1.49E5
*Nb 95 1.86E4 1.03E4 5.66E3 1. 00E4 7. 51E5 9. 68E4 Mo 99 1.69E2 3.22E1 4.11E2 1.54E5 2.69E5 I 131 3.54E4 4.91E4 2.64E4 1.46E7 8.40E4 6.49E3 I 133 1.22E4 2.05E4 6.22E3 2.92E6 3.59E4 1.03E4 Cs 134 5.02E5 1.13E6 5.49E5 3.75E5 1.46E5 9.76E3 Cs 137 6.70E5 8.48E5 3.11E5 3.04E5 1.21E5 8.48E3 Ba 140 5.47E4 6.70E1 3.52E3 2. 28E1 2. 03E6 2. 29E5
La 140 4.79E2 2.36E2 6.26E1 2. 14E5 4.87E5 Ce 141 2.84E4 1.90E4 2.17E3 8. 88E3 6. 14E5 1.26E5 Ce 144 4.89E6 2.02E6 2.62E5 1.21E6 1.34E7 8.64E5

Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-10 R i VALUES INHALATION ADULT m~rem/ r 3 pCi/m NUCLIDE BONE LIVER Y. BODY YHYROID KIDNEY LUNG GI-LLI H 3 1.26E3 1 .26E3 1.26E3 1.26E3 1 .26E3 1.26E3 C 14 1.82E4 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 3.41E3 Cr 51 1.00E2 5.95E1 2.28E1 1.44E4 3.32E3 Mn 54 3. 96E4 6.30E3 9. 84E3 1.40E6 7. 74E4 Fe 59 1.18E4 2.78E4 1.06E4 1.02E6 1-88E5 Co 58 1.58E3 2.07E3 9.28E5 1.06E5 Co 60 1.15E4 1.48E4 5.97E6 2.85E5 Zn 65 3.24E4 1.03E5 4.66E4 6.90E4 8.64E5 5.34E4 Sr 89 3.04E5 8.72E3 1.40E6 3.50E5 Sr 90 9.92E7 6.10E6 9.60E6 7.22E5 Zr 95 1.07E5 3.44E4 2.33E4 5.42E4 1.77E6 1.50E5
Nb 95 1.41E4 7.82E3 4.21E3 7.74E3 5.05E5 1.04E5 Mo 99 1.21E2 2.30E1 2.91E2 9. 12E4 2.48E5 I 131 2.52E4 3.58E4 2.05E4 1.19E7 6.13E4 6. 28E3 I 133 8.64E3 1.48E4 4.52E3 2.15E6 2.58E4 8.88E3 Cs 134 3.73E5 8.48E5 7.28E5 2.87E5 9.76E4 1.04E4 Cs 137 4.78E5 6.21E5 4.28E5 2.22E5 7.52E4 8.40E3 Ba 140 3.90E4 4.90E1 2.57E3 1.67E1 1.27E6 2.18E5

La 140 3.44E2 1.74E2 4.58El 1.36E5 4.58E5 Ce 141 1-99E4 1.35E4 1.53E3 6.26E3 3. 62E5 1.20E5 Ce 144 3.43E6 1.43E6 1.84E5 8.48E5 7.78E6 8.16E5

Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-11 R i VALUES GROUND PLANE ALL AGE GROUPS m 2 - mrem/yr 'pCi/se c NUCLIDE TOTAL BODY SKIN H 3 C 14 Cr 51 4.65E6 5.50E6 Mn 54 1.40E9 1.64E9 Fe 59 2.73E8 3.20E8 Co 58 3.80E8 4.45E8 Co 60 2. 15E10 2.53E10 Zn 65 7.46E8 8.57E8 Sr 89 2. 16E4 2.51E4 Sr 90 Zr'5 2.45E8 2.85E8
*Nb 95 2.50E8 2. 94E8 Mo 99 3.99E6 4.63E6 I 131 1.72E7 2.09E7 I 133 2.45E6 2.98E6 Cs 134 6. 83E9 7.97E9 Cs 137 1 .03E10 1.20E10 Ba 140 2.05E7 2.35E7 <<La 140 1.47E8 1.66E8 Ce 141 1.37E7 1.54E7 Ce 144 6.96E7 8.07E7
Daughter Decay Product. Activity level and effective half life assumed to equal. parent nuclide.
-7>- May 1986
TABLE 3-12 R i VALUES COW MILK INFANT 2 m harem/yr -'. uCi/sec NUCLIDE BONE LIVER Y. BODY YHYROYD KIDNEY LUNG GI-LLI
*H 3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 2.38E3 *C 14 3.23E6 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 6.89E5 Cr 51 8.35E4 5.45E4 1.19E4 1 .06E5 2.43E6 Mn 54 2.51E7 5.68E6 5.56E6 9.21E6 Fe 59 1.22ES 2.13E8 8.38E7 6.29E7 1.02E8 Co 58 1.39E7 3.46E7 3.46E7 Co 60 5.90E7 1.39ES 1.40ES Zn 65 3.53E9 1.21E10 5.58E9 5.87E9 1. 02E10 Sr 89 6.93E9 1.99E8 1.42ES Sr 90 8.19E10 2.09E10 1.02E9 Zr 95 3.85E3 9.39E2 6.66E2 1 .01E3 4.68E5
Nb 95 3. 93E5 1.62E5 9.35E4 1.16E5 1.37E8 Mo 99 1.04ES 2.03E7 1.55ES 3.43E7 I 131 1.36E9 1.60E9 7.04E8 5.26E11 1.87E9 5.72E7 I 133 1.81E7 2.64E7 7.72E6 4.79E9 3.10E7 4.46E6 Cs 134 2.41E10 4.49E10 4.54E9 1.16E10 4.74E9 1.22E8 Cs 137 3.47E10 4.06E10 2.88E9 1.09E10 4.41E9 1.27E8 Ba 140 1.21E8 1.21E5 6.22E6 2.87E4 7.42E4 2.97E7

La 140 1.86E2 7.35E1 1.89E1 8.63E5 Ce 141 2.28E4 1.39E4 1.64E3 4.28E3 7.18E6 Ce 144 1.49E6 6.10E5 8.34E4 2.46E5 8.54E7

mrem/yr per uCi/m
<<*Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. May 1986
TABLE 3-13 R i VALUES COW MILK - CHILD 2 m harem/yr -'. uCi/sec NUCLIDE BONE LIVER Y. BODY YHYROID KIDNEY LUNG GI-LLI
*H 3 1.57E3 1.57E3 1.57E3 1 .57E3 1.57E3 1 .57E3 *C 14 1.65E6 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 3.29E5 Cr 51 5.27E4 2.93E4 7.99E3 5.34E4 2.80E6 Mn 54 1.35E7 3.59E6 3. 78E6 1.13E7 Fe 59 6.52E7 1.06ES 5.26E7 3.06E7 1. 10E8 Co 58 6.94E6 2.13E7 4.05E7 Co 60 2.89E7 8.52E7 1.60E8 Zn 65 2.63E9 7.00E9 4.35E9 4.41E9 1.23E9 Sr 89 3.64E9 1.04E8 1.41ES Sr 90 7.53E10 1.91E10 1.01E9 Zr 95 2.17E3 4.77E2 4.25E2 6.83E2 4.98E5
Nb 95 2.10E5 8.19E4 5.85E4 7. 70E4 1.52ES Mo 99 4.07E7 1 .01E7 8.69E7 3.37E7 I 131 6.51ES 6.55ES 3.72E8 2. 17E11 1.08E9 5.83E7 I 133 8.58E6 1.06E7 4.01E6 1.97E9 1.77E7 4.27E6 Cs 134 1.50E10 2.45E10 5.18E9 7. 61E9 2. 73E9 1.32E8 Cs 137 2.17E10 2.08E10 3.07E9 6.78E9 2.44E9 1.30E8 Ba 140 5.87E7 5.14E4 3.43E6 1.67E4 3.07E4 2.97E7

"La 140 8.92E1 3.12E1 1.05El 8.69E5 Ce 141 1.15E4 5.73E3 8.51E2 2.51E3 7.15E6 Ce 144 1.04E6 3.26E5 5.55E4 1.80E5 8.49E7

mrem/yr per uCi/m
Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-14 Ri VALUES - COW MILK - TEEN 2 m harem/yr -'. uCi/sec NUCLIDE BONE LIVER Y. BODY IHYEOID KIDNEY LUNG GI-LLI
*H 3 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 9.94E2 *C 14 6.70E5 1.34E5 1.34E5 1.34E5 1.34E5 1-35E5 1-34E5 Cr 51 2.58E4 1.44E4 5.66E3 3.69E4 4.34E6 Mn 54 9.01E6 1.79E6 2.69E6 1.85E7 Fe 59 2.81E7 6.57E7 2.54E7 2.07E7 1.55ES Co 58 4.55E6 1.05E7 6. 27E7 Co 60 1.86E7 4.19E7 2.42E8 Zn 65 1.34E9 4.65E9 2.17E9 2.97E9 1.97E9 Sr 89 1.47E9 4.21E7 1.75ES Sr 90 4.45E10 1.10E10 1.25E9 Zr 95 9.34E2 2.95E2 2.03E2 4.33E2 6.80E5 **Nb 95 9.32E4 5.17E4 2.85E4 5.01E4 2.21ES Mo 99 2.24E7 4.27E6 5.12E7 4.01E7 I 131 2.68ES 3.76ES 2.02E8 1.10E11 6.47E8 7.44E7 I 133 3.53E6 5.99E6 1.83E6 8.36E8 1.05E7 4.53E6 Cs 134 6.49E9 1.53E10 7.08E9 4.85E9 1.85E9 1.90ES Cs 137 9.02E9 1.20E10 4.18E9 4.08E9 1.59E9 1.71ES Ba 140 2.43E7 2.98E4 1.57E6 1.01E4 2.00E4 3.75E7
La 140 3.73E1 1.83E1 4.87EO 1.05E6 Ce 141 4.67E3 3.12E3 3.58E2 1.47E3 8.91E6 Ce 144 4.22E5 1.74E5 2.27E4 1.04E5 1.06E8 "mrem/yr per uCi/m

Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-15 MILK ADULT R i VALUES COW 2 m ~rem/yr -'. uCi/sec NUCLIDE BONE LIVER 1. BODY YHYROID KIDNEY LUNG GI-LLI
*H 3 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 7.63E2 *C 14 3.63E5 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 7.26E4 Cr 51 1.48E4 8.85E3 3.26E3 1.96E4 3.72E6 Mn 54 5.41E6 1.03E6 1.61E6 1.66E7 Fe 59 1.61E7 3.79E7 1.45E7 1.06E7 1.26E8 Co 58 2.70E6 6.05E6 5.47E7 Co 60 1.10E7 2.42E7 2.06ES Zn 65 8.71ES 2.77E9 1.25E9 1.85E9 1.75E9 Sr 89 7.99E8 2.29E7 1.28ES Sr 90 3.15E10 7.74E9 9.11ES Zr 95 5.34E2 1.71E2 1.16E2 2.69E2 5.43E5
Nb 95 5.46E4 3.04E4 1.63E4 3. 00E4 1.84E8 Mo 99 1.24E7 2.36E6 2.81E7 2.87E7 I 131 1.48E8 2. 12E8 1. 21ES 6. 94E10 3. 63ES 5. 58E7 I 133 1.93E6 3.36E6 1.02E6 4.94E8 '5.86E6 3.02E6 Cs 134 3.74E9 8.89E9 7.27E9 2.88E9 9.55E8 1.56E8 Cs 137 4.97E9 6.80E9 4.46E9 2.31E9 7.68E8 1.32ES Ba 140 1.35E7 1.69E4 8.83E5 5.75E3 9.69E3 2.77E7
"*La 140 2.07E1 1.05E1 2.76EO 7.67E5 V Ce 141 2.54E3 1.72E3 1.95E2 7. 99E2 6.58E6 Ce 144 2.29E5 9.58E4 1.23E4 5.68E4 7.74E7 "mrem/yr per uCi/m "*Daughter Decay Product. -Activity level and effective half life assumed to equal parent nuclide. May 1986
TABLE 3-16 R VALUES COW MEAT - CHILD 2 m ~rem/yr -'. pCi/se c NUCLIDE BONE LIVER Y. BODY YHYROID KIDNEY LUNG GI-LLI "H 3 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2 2.34E2
*C 14 5.29E5 1.06E5 1.06E5 1.06E5 1.06E5 1.06E5 1.06E5 Cr 51 4.55E3 2.52E3 6.90E2 4.61E3 2.41E5 Mn 54 5.15E6 1.37E6 1.44E6 4.32E6 Fe 59 2.04ES 3.30E8 1.65ES 9.58E7 3.44ES Co 58 9.41E6 2. 88E7 5.49E7 Co 60 4.64E7 1.37E8 2.57E8 Zn 65 2.38E8 6.35ES 3.95E8 4.00ES 1.12ES Sr 89 2.65E8 7.57E6 1.03E7 Sr 90 7.01E9 1.78E9 9.44E7 Zr 95 1.51E6 3.32E5 2.95E5 4.75E5 3.46ES
Nb 95 2.41E6 9.38E5 6. 71E5 8.82E5 1.74E9 Mo 99 5.42E4 1.34E4 1.16E5 4.48E4 I 131 8.27E6 8.32E6 4.73E6 2.75E9 1.37E7 7.40E5 I 133 2.87E-1 3.55E-1 1 .34E-1 6.60E-1 5.92E-1 1.43E-1 Cs 134 6.09ES 1.00E9 2.11ES 3.10E8 1.11ES 5.39E6 Cs 137 8.99E8 8.60ES 1.27E8 2.80E8 1 .01E8 5.39E6 Ba 140 2.20E7 1.93E4 1.28E6 6.27E3 1.15E4 1.11E7
"*La 140 1.67E2 5.84E1 1.97E1 1.63E6 Ce 141 1.17E4 5.82E3 8.64E2 2.55E3 7.26E6 Ce 144 1.48E6 4.65E5 7.91E4 2.57E5 1.21E8
mrem/yr per uCi/m

~Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-17 R i VALUES COW MEAT - TEEN 2 m harem/y . pCi/sec NUCLIDE BONE LIVER Y. BODY YHYROYD KIDNEY LUNG GI-LLI
*H 3 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 1.94E2 *C 14 2.81E5 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 5.62E4 Cr 51 2.93E3 1.62E3 6.39E2 4.16E3 4.90E5 Mn 54 4.50E6 8.93E5 1.34E6 9.24E6 Fe 59 1.15ES 2.69ES 1 .04ES 8.47E7 6.36E8 Co 58 8.05E6 1.86E7 l. 11ES Co 60 3.90E7 8.80E7 5.09E8 Zn 65 1.59E8 5.52ES 2.57E8 3.53ES 2.34ES Sr 89 1.40E8 4.01E6 1.67E7 Sr 90 5.42E9 1.34E9 1.52E8 Zr 95 8.50E5 2.68E5 1 .84E5 3.94E5 6. 19E8
Nb 95 1.40E6 7.74E5 4.26E5 7. 51E5 3.31E9 Mo 99 3.90E4 7.43E3 8.92E4 6.98E4 I 131 4.46E6 6.24E6 3.35E6 1.82E9 1.07E7 1.23E6 I 133 1.55E-1 2.62E-l S.OOE-2 3.66El 4.60E-1 1.99E-1 Cs 134 3.46ES 8. 13E8 3.77ES 2. 58ES 9. 87E7 1. 01E7 Cs 137 4.88ES 6.49E8 2.26ES 2.21ES 8.58E7 9.24E6 Ba 140 1.19E7 1.46E4 7.68E5 4.95E3 9.81E3 1.84E7

"La 140 9.12E1 4.48El 1.19E1 2.57E6 Ce 141 6.19E3 4.14E3 4.75E2 1.95E3 1.18E7 Ce 144 7.87E5 3.26E5 4.23E4 1.94E5 1-98ES

mrem/yr per pCi/m3.
+*Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. May 1986
TABLE 3-18 ADULT R i VALUES 2 COW MEAT m ~rem/yr -.'uCi/sec NU CLIDE BONE jIVER T. BODY THYROID KIDNEY jUNG GI- jjI
H 3 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 AC 14 3.33E5 6. 66E4 6. 66E4 6. 66E4 6. 66E4 6. 66E4 6. 66E4 Cr 51 3.65E3 2.18E3 8.03E2 4.84E3 9.17E5 Mn 54 5.90E6 1.13E6 1.'76E6 1.81E7 Fe 59 1.44ES 3.39E8 1.30ES 9.46E7 1.13E9 Co 58 1.04E7 2.34E7 2.12ES Co 60 5.03E7 1.11E8 9.45ES Zn 65 2.26ES 7.19E8 3.25E8 4.81E8 ~
4. 53ES Sr 89 1.66E8 4.76E6 2.66E7 Sr 90 8.38E9 2.06E9 2.42E8 Zr 95 1.06E6 3.40E5 2.30E5 5.34E5 1.08E9
Nb 95 1.79E6 9.94E5 5.35E5 9.83E5 6.04E9 Mo 99 4.71E4 8.97E3 1.07E5 1.09E5 I 131 5.37E6 7.67E6 4.40E6 2.52E9 1.32E7 2.02E6 I 133 1.85E-l 3.22E-1 9.81E-2 4.73El 5.61E-1 2.89E-1 Cs 134 4.35E8 1.03E9 8.45E8 3.35ES l. 11E8 1. 81E7 Cs 137 5.88ES 8.04ES 5.26E8 2.73ES 9.07E7 1.56E7 Ba 140 1.44E7 1.81E4 9.44E5 6. 15E3 1.04E4 2.97E7
"*La 140 1.11E2 5.59El 1.48El 4. 10E6 Ce 141 7.38E3 4.99E3 5.66E2 2.32E3 1.91E7 Ce 144 9.33E5 3.90E5 5.01E4 2.31E5 3. 16E8 "mrem/yr per >iCi/m ~"Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. May 1986
TABLE 3-19 R VALUES VEGETATION CHILD 2 m ~rem/yr . pCi/sec NUCLIDE BONE I IVER 1. BODY THYROID KIDNEY LUNG G I-LLI ~H 3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 "C 14 3.50E6 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5 7.01E5 Cr 51 1.17E5 6.49E4 1.77E4 1.18E5 6.20E6 Mn 54 6.65E8 1.77E8 1.86ES 5.58E8 Fe 59 3.97E8 6.42E8 3.20E8 1.86E8 6.69E8 Co 58 6.45E7 1.97E8 3.76ES Co 60 3.78E8 1.12E9 2. 10E9 Zn 65 8.12E8 2.16E9 1.35E9 1.36E9 3. 80E8 Sr 89 3.59E10 1.03E9 1.39E9 Sr 90 1.24E12 3. 15E11 1.67E10 Zr 95 3.86E6 8.50E5 7.56E5 1.22E6 8.86E8
Nb 95 7.50E5 2.92E5 2.09E5 2.74E5 5.40E8 Mo 99 7.70E6 1.91E6 1.65E7 6.37E6 I 131 1.43ES 1.44E8 8.16E7 4.75E10 2.36E8 1.28E7 I 133 3.52E6 4.35E6 1.65E6 8.08E8 7.25E6 1.75E6 Cs 134 1.60E10 2.63E10 5.55E9 8. 15E9 2. 93E9 1.42ES Cs 137 2.39E10 2.29E10 3.38E9 7.46E9 2.68E9 1.43ES Ba 140 2.77ES 2.43E5 1.62E7 7. 90E4 1.45E5 1.40ES

La 140 3.37E4 1.18E4 3.97E3 3.28E8 Ce 141 6.56E5 3.27E5 4.85E4 1.43E5 4.08E8 Ce 144 1.27ES 3.98E7 6.78E6 2.21E7 1 .04 E10
<<mrem/yr per uCi/m ~ ""Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. May 1986
TABLE 3-20 R i VALUES - VEGETATION TEEN 2 m ~rem/yr -'. uCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI <<H 3 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3 2.59E3
*C 14 1.45E6 2.91E5 2. 91E5 2. 91E5 2. 91E5 2. 91E5 2. 91E5 Cr 51 6.16E4 3.42E4 1.35E4 8.79E4 1.03E7 Mn 54 4.54ES 9.01E7 1.36ES 9.32ES Fe 59 1.79ES 4.18ES 1.61E8 1.32E8 9.89ES 4.37E7 1.01E8 6. 02E8 Co 60 2.49ES 5.60E8 3.24E9 Zn 65 4.24E8 1.47E9 6.86E8 9.41ES 6.23ES Sr 89 1.51E10 4.33ES 1.80E9 Sr 90 7.51Ell 1.85E11 2.11E10 Zr 95 1.72E6 5.44E5 3.74E5 7.99E5 1.26E9
Nb 95 3.44E5 1.91E5 1.05E5 1.85E5 8. 16ES Mo 99 5.64E6 1.08E6 1.29E7 1 .01E7 I 131 7. 68E7 1.07E8 5. 78E7 3. 14E10 l. 85E8 2. 13E7 I 133 1.93E6 3.27E6 9.98E5 4.57ES 5.74E6 2.48E6 Cs 134 7.10E9 1.67E10 7.75E9 5.31E9 2.03E9 2.08E8 Cs 137 1.01E10 1.35E10 4.69E9 4.59E9 1.78E9 1.92ES Ba 140 1.38E8 1.69E5 8. 91E6 5. 74E4 1. 14E5 2. 13E8

La 140 1.69E4 8.32E3 2.21E3 4.78ES Ce 141 2.83E5 1.89E5 2.17E4 8.89E4 5.40E8 Ce 144 5.27E7 2.18E7 2.83E6 1.30E7 1.33E10
>mrem/yr per uCi/m <<<<Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. May 1986
TABLE 3-21 R i VALUES - VEGETATION ADULT 2 m ~rem/yr . pCi/sec NUCLIDE BONE LIVER T. BODY THYROID KIDNEY LUNG GI-LLI +H 3 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3 2.26E3
*C 14 8.97E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 1.79E5 Cr 51 4.64E4 2.77E4 1.02E4 6.15E4 1.17E7 Mn 54 3.13E8 5. 97E7 9.31E7 9.58E8 Pe 59 1.26E8 2.96E8 1.13ES 8.27E7 1.02E9 Co 58 3.08E7 6.90E7 6.24E8 Co 60 1.67ES 3.69ES 3.14E9 Zn 65 3.17ES 1.01E9 4.56E8 6. 75E8 6.36E8 Sr 89 9.96E9 2.86E8 1.60E9 Sr 90 6.05Ell 1.48E11 1. 75E10 Zr 95 1.18E6 3.77E5 2.55E5 5.92E5 1.20E9
"Nb 95 2.41E5 1.34E5 7.20E4 1.32E5 8.13ES Mo 99 6. 14E6 1.17E6 1.39E7 1.42E7 I 131 8.07E7 1.15E8 6.61E7 3.78E10 1.98E8 3.05E7 I 133 2.08E6 3.61E6 1.10E6 5.31ES 6.30E6 3.25E6 Cs 134 4.67E9 1.11E10 9.08E9 3.59E9 1.19E9 1.94ES Cs 137 6.36E9 8.70E9 5.70E9 2.95E9 9.81E8 1 .68E8 Ba 140 1.29E8 1.61E5 8.42E6 5.49E4 9.25E4 2.65ES
La 140 1.58E4 7.93E3 2.11E3 5.86E8 Ce 141 1.97E5 1.33E5 1.51E4 6.19E4 5. 09E8 Ce 144 3.29E7 1.38E7 1 .77E6 8.16E6 1.11E10

mrem/yr per uCi/m3
"*Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide. May 1986
TABLE 3-22 DISPERSION PARAMETERS AT CONTROLLING LOCATIONS* X/Q,Uv aad U VALUES VENT- DIRECTION ~l( / U~/(m E) Site Boundary"** 1,600 2.00 E-6 2.10E-9 Inhalation and Ground E (104') 1,800 1.42E-7 2. 90E-9 Plane Cow Milk ESE (130') 4,300 4. 11E-8 4. 73E-10 Goat Milk+" E (89') 12,500 1.75E-8 1.33E-10 Meat Animal E (114') 2,600 1.17E-7 1.86E-9 Vegetation E (96') 2,900 1.04E-7 1.50E-9 S EAUX Site Boundary*** 1,600 4.50E-8 6.00E-9 Inhalation and Ground E (109') 1,700 8. 48E-9 1.34E-9 Plane Cow Milk ESE (135') 49200 1.05E-8 3. 64E-10 Goat Milk** E (94') 12,500 1 .80E-8 1.84E-10 Meat Animal E (114') 2,500 1.13E-8 1.15E-9 Vegetation E (96') 2,800 1.38E-8 9.42E-10 NOTE: Inhalation and Ground Plane are annual average values. Others are grazing season only.
X/Q and D/Q values from NMP-2 ER-OLS..
X/Q and D/Q from C.T. Main Data Report dated November 1985.
y/Q and D/Q from NMP-2 FES, NUREG-1085, May 1985.
May 1986
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PARTICULATE IODINE NOBLE GAS COLLECTION COLLECTION MEASUREMENT STATION STATION STATION FILTER FILTER NEW NEW STACK DUALDILUTION PARTICU. IODINE lATE CARTRID-CARTRID. GES GES STACK ~ PUMP I I P V FRESHAIR FILTER P FLOW CONTROL DETECIOR DETECTOR DETECTOR (ISOKINETICI CHECK CHECK CHECK FLOW SENSORS AMPLIFIER SOURCE SOURCE SOURCE VALVES. ETC d ADC AMPLIFIER AMPLIFIERd ADC d ADC MULTICHANNEL lf'IDEO 44 TERMINAL INDUSTRIAL PROGRAMMABLE CONTROI.lER ANALYZER COMPUTER {MCA) PRINTER HOST COMPUTER DUAL DISK DRIVE FIGURE 3-6 BLOCK DIAGRAM TYPICAL GASEOUS EFFLUENT MONITORING SYSTEM NIAGARA MOHAWK POWER CORPORATION NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORT
4.0 URANIUM FUEL CYCLE The "Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows:
"Uranium fuel cycle means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light~atermooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power for public Use utilizing nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered non~ranium special nuclear and by-product materials from the cycle."
Section 3/4.11 .4 of the Technical Specifications requires that when the calculated doses associated with the effluent releases exceed twice the applicable quarter or annual limits, the licensee shall evaluate the calendar year doses and, if required, submit a Special Report to the NRC and limit subsequent releases such that the dose commitment to a real individual from all uranium fuel cycle sources is limited to 25 mrem to the total body or any organ (except the thyroid, which is limited to 75 mrem). This report is to demonstrate that radiation exposures to all real individuals from all uranium fuel cycle sources (including all liquid and gaseous effluent pathways and direct radiation) are less than the limits in 40 CFR Part 190. If releases that result in doses exceeding the 40 CFR 190 limits have occurred, then a variance from the NRC to permit such releases will be requested and reduce subsequent releases. if possible, action will be taken to The report to the NRC shall contain: Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site, that contribute to the annual dose of the maximum exposed member of the public.
2) Identification of the maximum exposed member of the public and a determination of the total annual dose to this person from all existing pathways and sources of radioactive effluents and direct radiation.
The total body and organ doses resulting from radioactive material in liquid effluents from Nine Mile Point Unit 2 will be summed with the doses resulting from the releases of noble gases, radioiodines, and particulates'he direct dose components will also be determined by either calculation or actual measurement. Actual measurements will utilize environmental TLD dosimetry. Calculated measurements will utilize engineering calculations to determine a projected direct dose components In the event calculations are used, the methodology will be detailed as required in Section 6.9.1.8 of the Technical Specifications. The doses from Nine Mile Point Unit 2 will be added to the doses to the maximum exposed individual that are contributed from other uranium fuel cycle operations within 5 miles of the site. May 1986
4.0 (Cont') For the purpose o f calculating doses, the results of the Environmental Monitoring Program may be included to provide more r ef ined estimates of doses to a real maximum exposed individual. Estimated doses, as calculated from station effluents, may be replaced by doses calculated from actual environmental sample results. 4,1 Evaluation of Doses From Liquid Effluents For the evaluation of doses to real members of the public from liquid effluents either calculational data based on liquid effluents and the equations found in section 2.3 of the ODCM may be used or actual fish sample and shoreline sediment sample data may be used. Doses calculated from actual sample data will only consider the fish and shoreline sediment pathways since other possible aquatic pathways are not considered significant. Doses to members of the public based on actual sample analysis data will be calculated using the equations and factors, as applicable, found in Regulatory Guide 1.109. The methodology used will be presented in detail as required by section 6.9.1 .8 of the Technical Specifications. Equations used to evaluate fish and shoreline sediment samples are based on Regulatory Guide 1.109 methodology'ecause of the sample medium type and the half-lives of the radionuclides historically observed, the decay corrected portions of the equations are does not reduce the conservatism of the calculated doses but deleted'his increases the simplicity from an evaluation point of view. The dose from fish sample media is calculated as.'1) Rwb Zi [Cif x u x 1000 x Diw Where: The total dose to the whole body of an adult in mrem per year Cif The concentration of radionuclide i in fish samples in pCi/gram The consumption rate of fish for an adult (21 kg per year) 1000 Grams per kilogram Diwb The dose factor for radionuclide adult (R.G. 1.109, Table E-11) i for the whole body of an The fractional portion of the year over which the dose is applicable. (2) Rl Zi [Cif x u x 1000 x Dil x f] May 1986
Where '. Rl The total dose to the liver of an adult (maximum exposed organ) in mrem per year Ci f The concentration o f radionuclide i in fish samples in pCi/gram u The consumption rate of fish for an adult (21 kg per year) 1000 ~ Grams per kilogram Dil The dose factor for radionuclide adult (R.G. Table E-ll) i for the liver of an f ~ The fractional portion of the year over which the dose is applicable. The dose from shoreline sediment sample media is calculated as: Rwb Zi [Cis x u x 40,000 x 0.3 x Diwb and Rsk Zi [Cis x u x 40,000 x 0.3 x Disk x Where: The total dose to the whole body of a teenager (maximum exposed age group) in mrem per year Rsk The total dose to the skin of a teenager (maximum exposed age group) in mrem per year Cis The concentration of radionuclide i in shorelin'e sedimen t in pCi/gram The usage factor. This is assumed as 67 hours per year by a teenager
40) 000 The product of the assumed density of shoreline sediment (40 kilogram per square meter to a depth of 2.5 cm) times the number of grams per kilogram 0.3 The shore width factor for a lake Diwb The dose factor for radionuclide 1.109, Table E-6) i for the total body (R.G.
Disk The dose factor for radionuclide 1 for the skin (R.G. 1.109, Table E-6) The fractional portion of the year over which the dose is applicable May 1986
4.2 Evaluation of Doses From Gaseous Effluents For the evaluation of doses to real members of the public from gaseous effluents, the pathways contained in section 3.0 of the ODCM will be considered and include ground deposition, inhalation, cows milk, goats milk, meat, and food products (vegetation) ~ However, any updated field data may be utilized that concerns locations of real individuals, real time meteorological data, location of critical receptors, etc. Data from the most recent census and sample location surveys should be utilized. Doses may also be calculated from actual environmental sample media, as available. Environmental sample media data such as TLD, air sample, milk sample and vegetable (food crop) sample data may be utilized in lieu of effluent calculational data. Doses to members of the public from the pathways contained in ODCM section 3.0 as a result of gaseous effluents will be calculated using the dose factors of Regulatory Guide 1.109 or the methodology of the ODCM, as applicable. Doses calculated from environmental sample media will utilize the methodologies found in Regulatory Guide 1.109. 4.3 Evaluation of Doses From Direct Radiation Section 3.11 .4.a of the Technical Specifications requires that the dose contribution as a result of direct radiation be considered when evaluating whether the dose limitations of 40 CFR 190 have been exceeded. Direct radiation doses as a result of the reactor, turbine and radwaste buildings and outside radioactive storage tanks (as applicable) may be evaluated by engineering calculations or by evaluating environmental TLD results at critical receptor locations, site boundary or other special interest locations. For the evaluation of direct radiation doses utilizing environmental TLDs > the critical receptor in question, such as the critical residence, etc., will be compared to the control locations'he comparison involves the difference in environmental TLD results between the receptor location and the average control location result-May 1986
4.4 Doses to Members of the Public Within the Site Boundary ~ Section 6.9. 1.8 of the Nine Mile Point Unit 2 Technical Specifications requires that the Semiannual Effluent Release Report include an assessment of the radiation doses from radioactive liquid and gaseous effluents to members of the public due to their activities inside the site boundary as defined by Figure 5.1.3 of the specifications. A member of the public, as defined by the Technical Specifications, would be represented by an individual who visits the sites'nergy Information Center for the purpose of observing the educational displays or for picnicing and associated activities. It is assumed that an individual would spend four hours per week for twelve weeks at the Energy Information Center. The time spent at the facility is assumed to occur from approximately July 1 to September 30 of each year. Thus, the first Semiannual Effluent Release Report will not address this particular dose because the summer season is the period of concern. The second report will address this dose based on forty eight hours occupancy. Other time periods of the year are not considered because any time spent inside the site boundary during months other than July-September is estimated to be less than 2-3 hours-The pathways considered for the evaluation include the inhalation pathway with the resultant lung dose and the direct radiation dose pathway with the associated total body dose. The direct radiation dose pathway, in actuality, includes several pathways. These include: the direct radiation gamma dose to an individual from on overhead plume, a submersion gamma plume dose, and a ground plane dose (deposition). Other pathways, such as the ingestion pathway, are not applicable'n addition, pathways associated with water related recreational activities are not applicable here. These inlcude swimming and wading which are prohibited at the facility. The inhalation pathway is evaluated by identifying the applicable radionuclides (radioiodine, tritium and particulates) in the effluent for the appropriate time period. The radionuclide concentrations are then multiplied by the appropriate X/9 value, inhalation dose factor, air intake rate, and the fractional portion of the year in question. Thus, the inhalation pathway is evaluated using the following equation adapted from Regulatory Guide 1.109. NOTE: The following equation is adapted from equations C-3 and C-4 of Regulatory Guide 1.109. Since many of the factors are in units of pCi/m , m /sec., etc., and since the radionuclide decay expressions have been deleted because of the short~ distance to the receptor location, the equation presented here is not identical to the Regulatory Guide equations. R ~ Z i[Ci F X/Q DFA i)a Ra t] May 1986
4' (Cont'd) where R the dose for the period in question to the lung ()) for all radionuclides (i) for the adult age group (a) in mrem per time period. Ci The concentration in the stack release of radionuclide in average pCi/m for the period in question F = Average effluent flowrate in m /sec . X/Q The plume dispersion parameters'or a location 0.50 miles west of NMP-1 are 2.88E-7 (X/Q) and 2.84E-9 (D/Q) for the NMP-2 Vent and 7.31E-7 (X/Q) and 4.37E-9 (D/Q) for the NMP-2 stack (data from C.T. Main Report dated ll/85). A X/Q value based on real time meteorology may also be utilized for the period in question) . DFAiga the inhalation dose factor for radionuclide i, lung g, and adult age group a in mrem per pCi found on the Table E-10 of Regulatory Guide 1.109. Ra annual air intake for individuals in age group a in M per year (this value is 8 000 m per year and was obtained from Table E-5 of Regulatory Guide 1.109). fractional portion of the year for which radionuclide was detected and for which a dose is to be calculated (equals 0.0055 years). The direct radiation gamma dose pathway includes any gamma doses from an overhead plume, submersion in the plume and ground plane dose (deposition). This general pathway will be evaluated by average environmental TLD readings't least two environmental TLD locations will be utilized and located in the approximate area of the Energy Information Center (EIC) and the facility picnic area. These TLDs will be placed in the field on approximately July 1 and removed on approximately September 30 of each year (this time interval is composed of one quarterly TLD collection period) ~ The average TLD readings will be ad)usted by the average control TLD readings'his is accomplished by subtracting the average quarterly control TLD value from the average EIC TLD value. The applicable quarterly control TLD values will be utilized after adjusting for the appropriate time period (as applicable). May 1986
5.0 ENVIRONMENTAL MONITORING PROGRAM 5.1 Sampling S tations The current sampling locations are specified in Table 5-1 and Figures 5.1-1, 5.1-2. The meteorological tower location is shown on Figure
5. 1-1. The location is shown as TLD location 817. The Environmental Monitoring Program is a )oint effort between the Niagara Mohawk Power Corporation and the New York Power Authority, the owners and operators of the Nine Mile Point Units 1 and 2 and the James A.FitzPatrick Nuclear Power Plants, respectively. Sampling locations are chosen on the basis of historical average dispersion or deposition parameters from both units. The environmental sampling location coordinates shown on Table 5-1 are based on the NMP-2 reactor centerline.
The average dispersion and deposition parameters for the two units have been calculated for a 5 year period, 1978 through 1982. These dispersion calculations are attached. The calculated dispersion or deposition parameters will be compared to the results of the annual land use census. If it is determined that a milk sampling location exists at a location that yields a significantly higher (e.g. 50X ) calculated D/Q rate, the new milk sampling location will be added to the monitoring program within 30 days. If a new location is added, the old location that yields the lowest calculated D/Q may be dropped from the program after October 31 of that year. 5.2 Interlaboratory Comparison Program Analyses shall be performed on samples containing known quantities of radioactive materials that are supplied as part of a Commission approved or sponsored Interlaboratory Comparison Program, such as the EPA Crosscheck Program. Participation shall be only for those media, e.g., air, milk, water, etc., that are included in the Nine Mile Point Environmental Monitoring Program and for which cross check samples are available. The Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission staff may evaluate the results'pecific sample media for which EPA Cross Check Program samples are available include the in air particulate filters following.'ross beta gamma emitters in air particulate filters I-131 in milk gamma emitters in milk gamma emitters in food product gamma emitters in water tritium in water I-131 in water May 1986
5.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements users by required the Technical Specifications are based on ANSI Standard N545> section 4.3. TLDs are defined as phosphors packaged for field In regard to the detection capabilities for thermoluminescent dosimeters, only one determination is required to evaluate the above capabilities per type of TLD. Furthermore, the above capabilities may be determined by the vendor who supplies the TLDs ~ Required detection capabilities are as follows. 5.3. 1 Uniformity shall be determined by giving TLDs from the same batch an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5% ~ A total of at least 5 TLDs shall be evaluated. 5.3.2 Reproducibility shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative standard deviations of the responses shall be less than 3.0X. A total of at least 4 TLDs shall be evaluated. 5.3.3 Dependence of exposure interpretation on the length of a field cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constants This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures'or these tests, the ratio of the response obtained in the field cycle to twice that obtained for half the field cycle shall not be less than 0.85. At least 6 TLDs shall be evaluated. 5.3.4 Energy dependence shall be evaluated by the response of TLDs to photons for several energies between approximately 30 keV and 3 MeV. The response shall not differ from that obtained with the calibration source by more than 25X for photons with energies greater than 80 keV and shall not be enhanced by more than a factor of two for photons with energies less than 80 keV. A total of at least 8 TLDs shall be evaluated . 5.3.5 The directional dependence of the TLD response shall be determined by comparing the response of the TLD exposed in the routine orientation with respect to the calibration source with the response obtained for different orientations. To accomplish this, the TLD shall be rotated through at least two perpendicular planes. The response averaged over all directions shall not differ from the response obtained in the standard calibration position by more than 10X. A total of at least 4 TLDs shall be evaluated. May 1986
Light dependence shall be determined by placing TLDs in the field for a period equal to the field cycle under the four conditions found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall not differ from those obtained for the TLDs wrapped in aluminum foil by more than 10' total of at least 4 TLDs shall be evaluated for each of the four conditions. Moisture dependence shall be determined by placing TLDs (that is, the phosphors packaged for field use) for a period equal to the field cycle in an area where the exposure rate is known to be constant. The TLDs shall be exposed under two conditions: (1) packaged in a thin, sealed plastic bag, and (2) packaged in a thin, sealed plastic bag with sufficient water to yield observable moisture throughout the field cycle The TLD or phosphor, as appropriate, shall be dried
~
before readout. The response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by more than 10X. A total of at least 4 TLDs shall be evaluated for each condition. Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is less than 10 uR/hr and the exposure during the field cycle is known. If necessary, corrections shall be applied for the dependence of exposure interpretation on the length of the field cycle (ANSI N545, section 4.3.3) ~ The average exposure inferred from the responses of the TLDs shall not differ from the known exposure by more than an exposure equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. A total of at least 3 TLDs shall be evaluated. May 1986
Nine Mile Point Nuclear Station Radiological Environmental Monitoring Program Sampling Locations Table 5.1 Type of "Map Sample Location Collection Site (Env. Program No.) Location Radioiodine and Nine Mile Point Road 1.8 mi 8 88' Particulates (air) north (R-1) Radioiodine and Co. Rt. 29 & Lake Road 1.1 mi 8 104'SE Particulates (air) (R-2) Radioiodine and Co. Rt. 29 (R-3) 1.5 mi 8 132':SE Particulates (air) Radioiodine and Village of Lycoming, NY 1.8 mi 8 143'E Particulates (air) (R-4) Radioiodine and Montario Point Road 16.4 mi 8 42'E Particulates (air) (R-5) Direct Radiation (TLD) North Shoreline Area (75) 0.1 mi 8 5' e Direct Radiation Direct Radiation (TLD) (TLD) North Shoreline Area (76) North Shoreline Area (77) O.l mi 0.2 mi 8 25'NNE 8 45'E Direct Radiation (TLD) 9 North Shoreline Area (23) 0.8 mi 8 70'NE Direct Radiation (TLD) 10 JAF east boundary (78) 1.0 mi 8 90' Direct Radiation (TLD) ll Rt ~ 29 (79) 1.1 mi 8 115'SE Direct Radiation (TLD) 12 Rt.,29 (80) 1.4 mi 8 133'E Direct Radiation (TLD) 13 Miner Road (81) 1.6 mi 8 159'SE Direct Radiation (TLD) 14 Miner Road (82) 1.6 mi 8 181' Direct Radiation (TLD) 15 Lakeview Road (83) 1.2 mi 8 200 .SSW Direct Radiation (TLD) 16 Lakeview Road (84) 1.1 mi 8 225',SW Direct Radiation (TLD) 17 Site Meteorological Tower (7) 0.7 mi 8 250'SW Direct Radiation (TLD) 18 Energy Information Center (18) 0.4 mi 8 265' "Map - See Figures 5.1-1 and 5.1-2 May 1986
Nine Mile Point Nuclear Station Unit 1 Radiological Environmental Monitoring Program Sampling Locations Table 5.1 ( Continued ) Type of
Map Sample Location Collection Site (Env. Pro ram No.) Location Direct. Radiation (TLD) 19 North Shoreline (85) 0.2 mi 8 294'N4 Direct Radiation (TLD) 20 North Shoreline (86) 0."1 mi 8 315'W Direct Radiation (TLD) 21 North Shoreline (87) 0.1 mi 8 341'%
Direct Radiation (TLD) 22 Hickory Grove Road (88) 4.5 mi 8 97' Direct Radiation (TLD) 23 Leavitt Road (89) 4.1 mi 8 ill'SE Direct Radiation (TLD) 24 Rt. 104 (90) 4.2 mi 8 135'E Direct Radiation (TLD) 25 Rt. 51A (91) 4.8 mi 8 156'SE Direct Radiation (TLD) 26 Maiden lane Road (92) 4.4 mi 8 183' Direct Radiation (TLD) 27 Co. Rt. 53 (93) 4.4 mi 8 205'Sb Direct Radiation (TLD) 28 Co ~ Rt ~ 1 (94) 4.7 mi 8 223'W Direct Radiation (TLD) 29 Lake Shoreline (95) 4.1 mi 9 237'Sb Direct Radiation (TLD) 30 Phoenix, NY Control (49) 19.8 mi 8 170' Direct Radiation (TLD) 31 S.W. Oswego, Control (14) 12.6 mi 8 226'h Direct Radiation (TLD) 32 Scriba, NY (96) 3.6 mi 8 199'SW Direct Radiation (TLD) 33 Alcan Aluminum, Rt. 1A (58) 3.1 mi 8 220'W Direct Radiation (TLD) 34 Lycoming, NY (97) 1.8 mi 8 143'E Direct Radiation (TLD) 35 New Haven, NY (56) 5.3 mi 8 123'SE Direct Radiation (TLD) 36 W. Boundary, Bible Camp (15) 0.9 mi 8 237'SW Direct Radiation (TLD) 37 Lake Road (98) 1.2 mi 8 101' Surface Water 38 OSS Inlet Canal (NA) 7.6 mi 8 235'W Surface Water 39 JAFNPP Inlet Canal (NA) 0.5 mi 8 70'NE .*Map See Figures 5. 1-1 and 5.1-2 (NA) - Not applicable May 1986
Nine Mile Point Nuclear Station Unit 1 Radiological Environmental Monitoring Program Sampling Locations Table 5.1 (Continued) Type of a'Map Sample Location Collection Site Location Shoreline Sediment 40 Sunset Bay Shoreline 1.5 mi 8 80' Fish 41 NMP Site Discharge Area 0.3 mi 8 315'W and/or Fish 42 NMP Site Discharge Area 0.6 mi 8 55'E Pish 43 Oswego Harbor Area 6.2 mi 8 235'W. Milk 44 Milk Location 850 9.3 mi 8 93' Milk 45 Milk Location 87 5.5 mi 8 107'SE Milk 46 Milk Location 816 5.9 mi 8 190' Milk 47 Milk Location 840 15.0 mi 8 223'W Food Product 48 Produce Location 86"" 1.9 mi 8 143'E (Bergenstock) Pood Product 49 Produce Location 81** 1.8 mi 8 96' (J. Parkhurst) Food Product 50 Produce Location b'2~* 1.9 mi 8 101' (Fox) Food Product Produce Location 85** 1.5 mi 8 114'SE (C. ST Parkhurst) Food Product Produce Location /13** 2.3 mi 8 122'SE (T. Parkhurst) Food Product 53 Produce Location 84"* 2.2 mi 8 123'SE (C. Lawton) Pood Product 54 Produce Location 87"* 15.0 mi 8 223'W (Mc Millen) Food Product .55 Produce Location 88** 12.6 mi 8 225'W (Denman)
" See Figures 5. 1-1 and 5. 1-2
Map Pood Product samples need not necessarily be colle'cted from all listed locations'ollected samples will be of the highest calculated site average D/Q.
-100- May 1986 >>>> Pll I" %W I g'>>>>, + g% A, 'bg>>'
6.0 DISCUSSION OF TECHNICAL SPECIFICATION REFERENCES S ection 3. 12. 1 o f the Technical Specifications, Table 3. 12-1 (Radiological Environmental Monitoring Program) references several footnotes to discussions in the ODCM. The following ODCM discussions are an attempt, on the part of the Commission and the licensee, to further clarify several of the requirements of Table 3.12-1. 6.1 Table 3.12-1, Footnote g f Representative composite sample aliquots ar e obtained rom sampling equipment that will obtain sample aliquots over short intervals ~ An example of a short interval is once per hour. Intervals of less than one hour are also acceptable'n addition, in order to be representative, the aliquot volume must be consistent over the required composite periods Sub&ntervals may be designed for sample collection as long as each sub-interval's contribution to the final composite volume is proportional to the duration of the sub-interval. For example, a monthly composite may consist of equal contributions from four weekly sub&ntervals, plus a contribution 3/7 of that volume from a fifth weekly sub-interval, to be representative of the monthly composite period. 6.2 Table 3.12-1, Footnote h Ground water in the vicinity of the site is not currently a drinking water pathway- The hydraulic gradient and recharge properties in the vicinity of the site currently cause ground water to flow in a northerly direction to lake Ontario ~ The results of such hydraulic gradient and recharge property studies are documented in the NMP-2 FSAR. Thus, any ground water utilized for drinking water or irrigation purposes is not affected by the site and therefore sampling of ground water is not currently required. In the event of significant seismic activity, however, the hydraulic gradient and recharge properties in the vicinity of the site may change. In this case it is possible that ground water utilized for drinking water or irrigation purposes may have a potential to become contaminated. Thus, in the event of a significant seismic occurrence, samples from one or two sources will be obtained as noted in Table 3.12-1, Section 3.b of the Technical Specifications until hydraulic investigations conclude that the previous hydraulic gradient and recharge property studies are unchanged. Investigations that conclude that the hydraulic gradient and recharge properties have changed and that there is a potential for contamination of ground water used for drinking water and/or irrigation will result in continuing any applicable ground water sampling.
-101- May 1986
0 6.3 Table 3.12-1, footnote i Currently, there are no drinking water sources (from Lake Ontario) that can be affected by the site under normal operating conditions . The closest drinking water source is near the City of Oswego. This source is located in an "upmurrent" direction for the majority of the time based on local Lake Ontario currents. In addition, the source is significantly affected by the "plume" from the Oswego River which enters Lake Ontario at a point between the site and the sources The source is located approximately eight miles to the west of the site. Other drinking water sources within 50 miles of the site range from 20 to 50 miles'hese sources are beyond any significant influence of the site. In the event a drinking water source (other than the source near the City of Oswego) is established within 10 miles of the site (current miles in contrast to air miles), then the new source will be evaluated for any significant dose effects based on dilution criteria. Sources found to be significantly affected by the site will be added to the Radiological Environmental Monitoring Program required by Table 3.12-1, section 3.C of the Technical Specifications. 6.4 Table 3.12-1, footnote 1 Considering the shoreline topography and land development within 10 miles of the site> and the dilution factors beyond 10 miles, only major irrigation projects where food products are irrigated with Lake Ontario water need be considered for specification 4.C of Table
3. 12-1 ~
Major irrigation projects are defined as agricultural projects where food products for human consumption are grown and irrigation water from Lake Ontario is used frequently." Major irrigation projects are not considered to be small private gardens located on the lake shore at summer residences or year-round residences where occasional use of lake water during times of draught has been observed ~ Major projects include pumps and piping systems, either permanent or temporary, that supply lake water to agricultural projects on a frequent use of lake water is not considered to have a basis'n-frequent significant effect on food products'herefore, such a situation does not constitute a major irrigation project. Currently, no major irrigation projects exist within 10 miles of the site (May 1986).
-102- May 1986
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NOTES TO FIGURE 5.1.3-1 (a) NMP1 Stack (height is 350') (b) NMP2 Stack (height is 430') (c) JAFNPP Stack (height is 385') (d) NMP1 Radioactive Liquid Discharge (Lake Ontario, bottom) (e) NMP2 Radioactive Liquid Discharge (Lake Ontario, bottom) (f) JAFNPP Radioactive Liquid Discharge (Lake Ontario, bottom) (g) Site Boundary (h) Lake Ontario-Shoreline (i) Meteorological Tower (j) Training Center (k) Energy Information Center Additional Information: - NMP2 Reactor Building Vent is located 187 feet above ground level - JAFNPP Reactor and Turbine Building Vents are located 173 feet above ground level - JAFNPP Radwaste Building Vent is 112 feet above ground level - The Energy Information Center and adjoining picnic area are UNRESTRICTED AREAS within the SITE BOUNDARY that are accessible to MEMBERS OF THE PUBLIC - Lake Road, a private road, is an UNRESTRICTED AREA within the SITE BOUNDARY accessible to MEMBERS OF THE PUBLIC NINE MILE POINT UNIT 2
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