ML18038A272
| ML18038A272 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 05/21/1987 |
| From: | Duell J, Leach E, Stuart C NIAGARA MOHAWK POWER CORP. |
| To: | |
| Shared Package | |
| ML17055D180 | List: |
| References | |
| PROC-870521, NUDOCS 8709030022 | |
| Download: ML18038A272 (142) | |
Text
NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 2 OFF-SITE DOSE CALCULATION MANUAL ODCM DATE AND INITIALS APPROVALS SIGNATURES REVISION 3 REVISION 4 REVISION 5 Radiation Protection Manager E. W. Leach Chemistry & Radiochemistry Supervisor J. N. Dnell
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Chentstry & Radiation Hanade t Q~Q ~/: ~
Superintendent C. L..Stuart ~ kl Station Superintendent NMPNS Unit 2 R. B. Abbott Superintendent s'eneral Nuclear Generation <<(g(Pj T. J. Perkins Summa of Pa es Revision 3 (Effective ~~19~87 )
PAGE DATE i-iii,1-4,6,8,10-14 17'8)20 53 55 89)91)92)94t 96-101 Say 1986 5,90, 93.-95,102 August 1986 9 October 1986 15,16 May 1987 54 Nay 1937 (.C.'I-l) 19 June 1987 (TCN-2) 7 June 1987 (TCN-3)
NIAGARA MOHAWK POWER CORPORATION THIS PROCEDURE NOT TO BE USED AFTER Hay 1989 SUBJECT TO PERIODIC REVIEW.
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NINE MILE POINT NUCLEAR STATION NINE MILE POINT UNIT 2 OFF-SITE DOSE CALCULATION=MANUAL (ODCM)
Continuation Cover Sheet Page 2 Summar . of Pa es (Cont'd)
PAGE DATE 36 September 1987 (TCN-4)
'V
'TABLE OF CONTENTS SECTION SUBJECT TS SECTION FAG 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 Deriv'ation 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 Pactor 19 Figure 2-1 Liquid Radwaste Treatment System 3.11.1.3 20-27 thru 2-8 Flow Diagrams 3. 11.3 Figure 2-9 Liquid Radiation Monitoring .28 Pigure 2-10 Off-line Liquid Monitor 29 3.0 GASEOUS EPPLUENTS 30 3~ 1 Gaseous Effluent Monitor Alarm Setpoints 30 3.1. 1 Basis 3.11.2.1 30
- 3. 1.2 Setpoint Determination Methodology 3.3.7.11 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.l.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
TABLE OF CONTENTS SECTION SUBJECT TS SECTION PAG E or TABLE or TABLE
~3 Organ Dose Rate Due to I-1 1, I-1 Tritium and Particulates with half" 3. 11.2. 1. b 39~1 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- Iodine, Particulate and Tritium 47-51 Organ Dose Rate Factors 3.4.4 Ri- Iodine, Particulate and Tritium 51-5 7 Organ Dose Fact'ors 3.4.5 X/Q and Wv- Dispersion Factors for Dose Rate 58
'.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 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 CONTEifTS S ECTION SUBJECT TS SECTION PAGE or TABLE or TABLE Table 3-22 X/Q, Wv and Ws- Dispersion 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 Dose's 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.I 98-'100 Program Sampling Locations 4.12.1 Table 3. 12-1 Note (a)
S 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 Table 3.12-1 note i 102 102 6.4 Table 3.12-1 note 1 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 and 4.11.1.1 .2 respectively .
Additionally, the method used to calculate the cumulative dose Contributions from liquid effluents and the methods used to assureb' thorough mixing and sampling. of liquid radioactive waste tanks to 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'resented.
Section 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
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 area. 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 f or 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 z< [0.8<<(F/f)*EL(Ci*CFi)]/[EL(Ci/MPCi)] + Background.
Where:
Alarm Setpoint The Radiation Detector Alarm Setpoint, cpm 0.8 Safety Factor, unitless F Rl Nonradioactive dilution flow rate, gpm. Service May 1986
Mater Flow ranges from 30,000 to 58,000 gpm.
Blowdown flow is typically 10,200 gpm.
Ci Concentration of isotope i in Radwaste tank prior to dilution, uCi/ml CFi Detector response for isotope i, net cpm/~iCi/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 i limit for isotope from 10CFR20 Appendix B, Table II, Column 2> uCi/ml Background 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/MP Ci ) The total fraction of the 10CFR20, Appendix B, Table II, Column 2 limit that's 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 Tem ering)
(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.
May 1986
2.1.2.2 Contaminated Dilution Water Radwaste Effluent Monitor Alarm Setpoin t Calcula tion:
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 B lowdown) 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 FMP C+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 Alarm 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/ml CFi Detector response for istope i, net cpm/uCi/ml See Table 2-1 for a list of nominal values MP Ci Concentration limit for isotope i from 10CFR20 Appendix B, Table II, Column 2, pCi/ml f The permissible Radwaste Effluent Flow rate, gpm Background 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 Zs[Fs+Cis] The total activity of nuclide i in all Effluent streams, pCi.-gpm/ml Zs [Fs] The total Liquid Effluent Flow rate, gpm (Service Water & CT Blowdown & Radwaste)
May 1986
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~Zi (Ci*CFi)]/[Zi (Ci/MPCi)] + Background.
Where:
Alarm Setpoint The Radiation Detector Alarm Setpoint, .
cpm 0.8 Safety Factor, unitless Ci Concentration of isotope containment, uCi/ml i in potential CFi Detector response for isotope i, net Table 2-1 for a liat of nominal values cpm/uCi/ml'ee MPCi Concentration limit for isotope i from 10CFR20
- Appendix B, 'Tabl'e II, Column 2, uCi/ml Background '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 potential contaminant, cpm Zi(Ci/MPCi) The total fraction of the 10CFR20, Appendix B, Table II, Column 2 limit that is in the potential containment, unitless..
CR+ZiCi An approximation to Zi(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 CFi.
2.1.3 Discussion August 1986
2.1.3.1 Liquid Radwaste Effluent Monitor The Liquid Radi'oactive Waste System Tanks are pumped to the dischar 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 enoughnearto field meet thermal pollution limits. However, it also assists in the dilution of any activity released. Service 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/pCi/ml, will be developed by dividing the noted detector response, Ei(Ci*CPi) 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:
a.(Ci CPi) - CR* zt.(ci)
Where the factors are all as defined above.
Por 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 PSAR Table 11 .2.5.
Pe-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 rate" oC Lhe t.iquid RadioacLive Wa"te System Tank" discharged is 165 gpm while dilution Flow from the minim<<m <<umber of Service Water Pumps alway" in service is over 30,000 gpm, and Cooling, T<>wer Rlowdown 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 concenLration actually. in the tank.
Potentially a discharge could continue even if the dist<ihution 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 err<><<ous 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 355 cpm at>ove background will be used unLil grab sample analysis with tl<~ <<<<iuired LID sensitiviLy on TS Table 4.11-L detects activity at such a levol that it ca<<not be discharged with the nominal TCN 3 setpoint. 355 cpm is the same nominal setpoint as l'or Lhe service water and cooling tower blowndown radiation monitors. These are all identical detectors.
June 1987
A sample calculation is presented below assuming tank concentrations equivalent to the diluted concentration presented in FSAR Table 11 .2.5 which is the expected, concentration of effluent waste after dilution that 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/gCi/ml cpm 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-8 1E-4 2.4E-4 8.42E7 1.98E+0 MN56 3.2E-7 IE-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 1.14E8 7. 8E+0 C060 1.4E-7 3E-5 4.7E-3 1.65E8 2.4E+1 NI63 3. 5E-10 'E-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 lE-4 6.8E-4 BR83 3.3E-8 3E-6 1.1E-2 BR84 8.9E-14 1.12E8 1.0E-5 SR89 3.6E-8 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 ~ 22E8 5.7E+1 SR92 7.6E-8 6E-5 1.2E-3 8.17E7 6. lE+0 Y91 1.7E-8 3E-5 5 'E-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 5o 2E-5 NB95 2.7E-9 1E-4 2.7E-5 8.5E7 2.4E-1 M099 6. OE-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 8E-5 8.5E-5 RU105 6.8E-8 lE-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-8 2E-5 7.4E-5 TE131M 2.4E-8 4E-5 6.0E-4 May 1986
ISOTOPE ACTIVITY MPC FRACTION DETECTOR CPM CONCENTRATION OF MPC RESPONSE TOTAL NAME uCi/ml >iCi/ma B/C cpm/>@i/ml cpm 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 lE-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 1.0E02 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 1E-4 2.3E-2 TOTALS 2.1E+1 ~ ~ +
For the example tank, per'missible discharge flow to ensure a concentration less than MPC in the discharge canal would be:
f m 10,200 " 1 ~ 324 gpm 2.1E1
- 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*Ei(Ci*CFi)]/[Ei(Ci/MPCi)]+ Background.
Where the Alarm Setpoint is in cpm, F is 10,200 gpm, Ei(Ci*CFi) is 2.4E+3 cpm, f is 165 gpm and Ei(Ci/MPCi) is 2.1E+1 unitless. These values yield an Alarm Setpoint of 5.7E+3 cpm above background, while the expected detector respmse 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 if grab sample analysis of the tank indicates that no activity detectable above the LLD requirements of Table 4.11-1 then the Liquid Radwaste Effluent Radiation Monitor Alarm Setpoint will be set at less than 355 cpm above Background, cpm. This is the same as the service water monitors initial alarm setpoint, see section 2.1.3.2.
October 1986
2.1 .3.2 Service Water and Cooling Tower Blowdown Effluent Monito r Service Water A and B and the Cooling Tower Blowdown are pumped to the discharge tunnel which in turn flows directly to Lake Ontario'ormal 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 Ki(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 it represents the most likely contaminate of Station Waters.
A two hour decay is chosen by )udgement 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 purpose's of documenting the calculation. It will be recalcul'ated 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'ETECTOR CPM NAME ACTIVITY OF MPC RESPONSE TOTAL CONCENTRATION uci/ml B/C 'pm/uci/ml cpm uci/ml A B C D 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 1E-4 4.0E-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 i<Ci/ml B/C cpm/.>Ci/ml cpm iiCi/ml B D E F (Ci) (MPCi) ( Ci/MPCi) (CFi) ( CiCFi )
NI65 3.0E-4 m-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. 1E-3 3E-6 1.0E3 7. 8E3 2.4E1 SR90 2.3E-4 3E-7 7.7E2 SR91 6.0E-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 2.47E8 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-l 8.35E7 3.3E 3 ZR97 2.9E-5 2E-5 1.5 NB95 4.1E-5 IE-4 4o lE-1 8.5E7 3.5E3 MO99 2 2E-2 4E-5 5.5E-1 2.32E7 5. 1E5 TC99M 2.2E-l 3E-3 7.3E1 2.32E7 5. IE6 RU103 5,4E-5 '8E-5 . 6.8E-1 RU105 4.5E-,3 1E-4 4.5E1 RU106 8. 4E-6
'E-5 8 'E-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 1.12E8 5.4E6 I131 1.3E-2 3E-7 4.3E4 1.01E8 1.3E6'.2E7 I132 1.2E>>l 8E-6 1.5E4 2.63E8 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-l 4E-6 3.584 1.17E8 1.6E7 CS 134 1.6E-4 9E-6 1.8E1 1.97E8 3.2E4 CS136 1.1E-4 6E-5 1.8 2.89E8 3.2E4 CS 137 2.4E-4 2E-5 1 ~ 2E1 7 '2E7 1.8E4 CS138 1.4E-2 1.45E8 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 1E-5 3.5 1.03E7 3.6E2 PR143 5E-5 ND147 6E-5 W187 6E-5 NP239 2 ~ 3E-1 lE-4 2.3E3 TOTALS ~ E Msy 1986
The Service Water Effluent Radiation Monitor Alarm Setpoint equation is.'larm Setpoint [0.8~Zi(Ci*CFi)]/[Q.(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 conservatives 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'or 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 zi [ Zs(Cis~Fs)/(MPCi*Zs(Fs))]
Where: MPC Fraction The limiting concentration of 10 CFR 20, Appendix B, Table II, Column 2, for or radionuclides other than dissolved entrained noble gases'or noble gases, the concentration shall be limited to 2 x 10E-4 microcurie/ml total activity, unitless Cis The concentration of nuclide i in particular effluent stream s, pCi/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),
uCi/ml Es(Cis*Fs) The total activity rate of nuclide i, in all the effluent streams s, pCi/ml *gpm Es(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.
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 dose 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 radioactivity 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 Zi [Ait>>XL*(dT1+Cil*F1)]
Where:
The cumulative dose commitment to the total body or any organ, t from the liquid effluents for the total time period '-'ZL(dTl) >
mrem dTl The length of the 1 th time period over which Cil and Fl are averaged for all liquid releases., hours Cil The average concentration of radionuclide, i, in undiluted liquid liquid effluents during time period dT1 from any re3.ease, 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.
Fl 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 l.
<3- 'ay 1986
Example Calculation Thyroid A sample of a radwaste tank indicates I-.131 and H-3 concentrations 1.5E-6 and 8.9E-3 pCi/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+El(dT1*Cil*F1)].
Where Dt mrem is the dose to organ t, Ait mrem/hr per pCi/ml is the ingestion dose commitment factor, dT hours is the time interval over which the release occurs, Ci pCi/ml is the undiluted concentration of nuclide i in the release and Fl unitless is the dilution factor for the release.
From Table 2-2 Ait is 7.21E4 and 3.37E-1 mrem/hr per pCl/ml respectively for I-131 and H-3 dose to the thyroid. From the discharge and dilution flow rate, Fl unitless can be calculated:
Fl 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 3.11.1.3.
~
2.4 Liquid Effluent Dose Factor Derivation Ait Ait mrem/hr per gi/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-ll of Regulatory Guide 1.109. The dose factor equation for a fresh water site is:
Ait ~ Ko+(Uw/Dw + Uf*BFi)~DFi May 1986
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 pCi/ml Ko Is the unit conversion factor, 1.14E5=1X10E6pCi/yCi x 1E3 ml/kg : 8760 hr/hr Uw 730 kg/yr, adult water consumption Uf 21 kg/yr, 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-ll of RG 1.109.
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>BPi)>DPi.
Example Calculation I-131 Thyroid Dose Factor for exposure from Liquid Effluents:
-'or DPi ~ 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-ml 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 it will be calculated and included in a revision to the ODCM.
2.5 Sampling Representativeness This section covers TS Table 4.11.1-1 note b concerning thoroughly mixin 3 each batch of liquid radwaste prior to sampling.
There are four tanks in the radwaste system designed to be discharged to the discharge canal. These tanks are labeled 4A, 4B, 5A, and 5B.
Liquid Radwaste Tank 5A and 5B at Nine Mile Point Unit 2 contain a sparger spray ring which assist the mixing of the tank contents while it is being recirculated prior to sampling. This sparger effectively mixes the tank four times faster than simple recirculation.
May 1987
Liquid Radvast>> Tanx 4A and 43 contain a mixing r ixd out no spa."
iVo credit is tauten for tne mixing e ffee ts of the ring. iVorrna l recirculation flow is l50 gpm for tank SA and 5B, 110 gpm for tank 4A and 4B anile e ic'< tank contains un to 25,000 gallons althougn 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*M Wnere:
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.
Is the factor that takes into account the mixing of the sparger, unitless, four for tank 5A and B, one for tank 4A and B.
Additionally the Alert .Alarm setpoint of the Liquid Radwaste Effluent Raliation 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 par't 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. The length of tubing between the continuously flowing sample and the sample spigot 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-day period. Cumulative doses will be determined at least once per 31 days (as indicated in Section 2.3) and doses will also be projected fully utilized.
if the radwaste treatment systems are not being Pull 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 Pig.
2-2) and Radwaste Demin. (see Pig. 2-3)
')
High Conductivity (Floor Drains): Floor Drain Filter (see Fig.
2-5) or Waste Evaporator (see Pig. 2-6)
May 1987
J
- 3) Regenerant Waste: Regenerant Fvaporator (see Fig". 2-8)
NOTE: Regenerant Evaporator and Waste Evaporator may be used 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
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 0. 85 Mo 99 0.232 Tc 99m 0. 232 Te 132 l. 12 Ba 140 0.499 Ce 144 0.103 Br 84 l. 12 I 131 1.01 I 132 2. 63 I 133 0.96 7 I 134 2.32 I 135 1. 17
- l. 97 Cs 134 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 1. 14 Co 60 1.65 Values from SWEC purchase specification NMP2-P281F.
May 1986
A'-ie, 2-2 Ai. VALUES LIQUID mr em ml hr pCi NUCLIDE T BODY GI-TRACT BONE LIVER KIDNEY THYROID LUNG H 3 3.37E-l 3.37E-1 3.37E-1 3.37E-1 3.37E-1 3.37E-1 Cr 51 1.28 3.21E2 2.81E-l 7.63E-1 1.69 2.54E1 TCN-2 Cu 64 4.72 8.57E2 1.01E1 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.~4E2 3.60E3 2.24E4 Sr 90 1.$ 6E5 1.60E4 5.52E5 Zr 95 5.91E-2 2.77E2 2.72E<<1 8.74E-2 1.37E-1 3.52E3 1.10E2 1.40E2 TCN-2 Mn 56 1.96E1 Mo 99 2.05El 2.50E2 1.08E2 2.44E2 Na 24 4.09E2 4.09E2 4.09E2 4.09E2 4.09E2 4.09E2 4.09E2 I 131 1.26E2 5.80E1 1.54E2 2.20E2 3.77E2 7.21E4 TCN-2 Ni 65 7.53 4.18E2 1.27E2 1.65E1 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-1 9.22E-2 1.57E-1 Ce 141 2.48E-3 8.36E1 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 June 1987
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RBLILA 6>'I<v lAKE V'HR SKP PUMP HEAI EICHANGER E1A I FROM NIS S1$ lfM I OA CA'I I IO CAS SISIIN OA CAF E O23 8 I I RHR SERVICE MAIER IB) n, la CKS S'lSllw RHR HEAI EICHANGER I SIP PU'IIP I OA CAI 1 'RE I IlGA FROM RNR HEAIEI. EIA FROM CCP HEAEEK, EIA.IB.IC aA cAE E2 I SERVICE WATER
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CIRCULAllNG IAIER PUMPS Rf 205 1 i Acl laISCHARCE OA CAI SNUIOOKN) El ALL LIOUIO RAOBASIE EFFLUENI OISCHARGES SFC FILIERIOEMINS SIC HEAI EICMANGERS ODCH Fig. ')-'I SFC PUMPS Rf 142 I SPENF FUEL OA CAI I SFC SFSIEN POOL aA cAI EE OA CAF E FIGURE 11.5-8 LIQUID RADIATIONMONI TOI(II N . SHEET 2 OF 2 NIAGARA MOHAWK POWER CORP(alii. NINE MILE POINT-UNIT 2 FINAL SAFETY ANALYSIS REPORI
INTERCHANGEABLE PURGE OR TEST CONNECTION GRAB S SAMPLER I S (2) I Tl I (2) TS I (2) I S DATA LIQUID ACQUISITION DETECTOR UNIT L (DAU) SAMPLER I I (9) FI FS PI I PUMP NOTES: (1) GLOBE VALVE, ALL OTHER MANUALLY (6) TS-TEMPERATURE SWITCH OPERATED VALVES ARE BALL VALVES (7) CS-CHECK SOURCE (2) REQUIRED ONLY IF SAMPLE FLUID TEMPERATURE EXCEEDS SELLERS (8) Pl-PRESSURE INDICATOR DETECTOR TEMPERATURE REQUIREMENTS (9) FI-FLOW INDICATOR (3) ~ NORMALLYCLOSED (10) FS-FLOW SWITCH (4) ~ NORMALLYOPEN (11) DRAIN CONNECTION (5) Tl-TEMPERATURE INDICATION ODCM Fig. 2-10 FIGURE 11.5-3 OFF-LINE LIQUID MONITOR NIAGARA MOHAWK POWER CORPORATION NINE MlLE 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/Reactox Building vent. (See Figure 3.5) The stack effluent point includes Turbine Building ventilation,'ain condenser offgas (after chaxcoal 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 fxom radioactive materials released in gaseous effluents from the site to areas at or beyond the SITE BOUNDR (see Figure 5. 1.3-1) shall be limited to the following:
- a. For noble gases.'ess than or equal to 500 mrem/yr to the whole body and less than or equal to 3000 mrem/yr to the skin, an
- b. For iodine-131, for iodine-133, for tritium, and for all
~
radionuclides with half-lives greater than 8 days.'ess 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 microcuries/second during offgas syst: em operation. 3.1.2 Setpoint Determination Methodology The alarm setpoint for Gaseous Effluent Noble Gas Monitors are baseed on a d ose ra t e li it m of 500 mrem/yr to the Whole Body. These monitors are sensitive to only noble gases.. Because o f thiss it is considered impractical to base their alarm setpoints on organ dose rates due to iodines or particulates. Additionally skin dose rate is never significantly greater than the whole body dose xate. The alarm setpoint for the Offgas Noble Gas monitor is based on a limit of 350,000 pCi/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 Setpoint < ~Ei CE~VE Alarm Setpoint Is the alarm setpoint of the Stack Effluent Monitor, uCi/se c May 1986
0.8 Is a Safety Factor, unitless 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, iiCi/ml Is the Stack effluent flow rate,. ml/sec Vi 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 a cled cance e out of the above expression. The equation ratios the basis, R, to the actual dose rate from the effluent, F *Zi (Ci AVi), annd multiplies the unitless result by the actual effluent release rate, F"Xi(Ci). Since the Stack Effluent Monitor actually measures release rate in uCi/sec the detector response does not enter in. 3.1,2,2
~ ~ ~ 'ent Noble Gas Detector Alarm Setpoint Equation:
0.8*R"Ei (Ci) Alarm Setpoint < Where. Alarm Setpoint Is the alarm setpoint of the Vent Effluent Monitox', uCi/sec 0.8 Is a Safety Factor Is a value of 500 mrem/yr or less depending upon the dose rate from othex'elease points within the site such that the total rate corresponds to < 500 mrem/yr Is the concentration of nuclide i, uCi/ml Is the Vent effluent flow rate, ml/sec (X/g)v Is the highest annual average atmospheric dispersion coefficient at the'ite boundry as listed in the Final Environmental Statement, NUREG 1085, Table D-2, 2.0E-6 sec/m3 Ki 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 31 May 1986
z(ci) Is the total concentraton o" noble gas nucL'>i~=~ the Vent effluent, -Ci/ml Q. (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 The equation ratios the basis, canceled out of the above expression. 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*ZL(Ci). Since the Vent Effluent Monitor actually measures release rate in >>Ci/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"Ei(Ci+CFi) Alarm Setpoint < f Ei Ci + Background Where: Alarm 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 f 'uclide, i, in
's Ci the concentration o the f
Of gas, uCi/ml CFi Is the Detector response to nuclide i, net cpm/pCi/ml See Table 3-1 for a list of nominal values. See section 3.1.3.3 for discussion Is the f Of gas System Flow rate, CFM Background Is the detector response when its chamber is filled with nonradioactive air, cpm Ei ( CiC Fi) Is the summation of the product of the nuclide concentration and corresponding detector response, net cpm Ei(Ci) Is the summation of the concentration of nuclides in offgas, uCi/ml . 32 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 and before the charcoal decay beds. The offgas, after decay, is exhausted to the main stack. The system will automatically
'ecombiners isolate if its pretreatment radiation monitor detects levels of radiation above the alarm setpoint.
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 it is higher than any adjacent buildings it 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 Fitzpatri.ck 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 at Unit 2 that both release points could be involved. Thus the 'ccur 'actor 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 i's 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 iiCi 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 the most significant contaminate of gaseous activity 'in the plant. The following calculation will be used for the initial Alarm Setpoint. It will be recalculated if a significant release is encountered. In that case the actual distribution of noble gases will be used in the 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 nCi/ml mrem-sec mrem/ r yr-sCf ml sec B C . Dsm(BaC) (Ci) (Vi) (Ci*Vi) KR83 8.74E-2 KR85 4.90E-4 3.28E-5 l. 61E-8 5.01E-3 KR85M 1.56E-1 3.21E-3 KR87 5. 23E-1 9.98E-3 5 '2E-3 KR88 5.32E-1 2 21E-2 l. 18E-2 KR89 1.63 1.92E-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 6.08E-3
- XE137 11 2.88E-3 XE138 1.93 1 33E-2 m 2.57E-2 AR41 1.61E-2 OT S ~ 28E-2 The alarm setpoint equation is:
Alarm Setpoint ~ 0.8*R+Zi(Ci)/Zi(Ci"Vi). Where the Alarm Setpoint is in IICi/sec, R is taken as 250mrem/yr, Z(Ci) is 8.36 IICi/ml and Z(Ci"Vi) is 9.28E-2 mrem/yr per m1/sec. These values yield an alarm setpoint of 1.80E4 IICi/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
'adiation. However, because it is a computer based multichannel analysis system it is able to accurately quantify the activity released in terms of (<Ci 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 it represents the most significant contaminate of gaseous activity in.
the plant. The following calculation will be used for the initial Alarm Setpoint. It will be recalculated if a significant release is encountered'n that case the actual distribution of noble gases will be used in the calculation. ISOTOPE ACTIVITY IMMERSION IMMERSIO N NAME CONCENTRATION FACTOR FACTOR i.Ci/ml mr em~3 mr em~3 yr-eCi yr ml C D~(B<<C) KR83 8. 74E-2 7.56E-2 6. 61E-3 KR85 4.90E-4 1.61E-1 7.90E-3 ZR85M 1. 56E-l 1.17E-3 1.82E2 KR87 5. 23E-1 5.92E3 3.10E3 KR88 5. 32E-l 1.47E4 7.82E3 KR89 1.63 1.66E4 2.71E4 KR90 1.56E4 XE131M 3.82E-4 9.15E1 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. 12E The Vent Effluent Noble Gas Monitor Alarm Setpoint equation is: Alarm Setpoint 0.&R*H.(Ci)/[(X/Q)v*8.(Ci*Ki)] ~ Where the Alarm Setpoint is in uCi/sec, R is 250mrem/yr, Q.(Ci) is 8.36 uCi/ml, (X/Q) is 2.0E-6 se c/m3 and Ei(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 it is sensitive to both gamma and beta radiation. Detector response 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 ACTIV1TY DETECTOR DETECTOR NAME CONCENTRATION RESPONSE CPM yCi/ml cpm/pCi/ml cpm B C D Ci) (CFi) (Ci"CFi KR83 8.74E-2 KR85 4 '0E-4 4 '083 F 11 KRBSM 1.56E-1 4.80E3' 7.5083 KR87 5.23E-1 '083 4.1883 KR88 5;32E-1 7.60E3 4 '4E3 KR89 1.63 KR90 XE131M 3.82E-4 XE133 2.068-1 1.7583 3.60E2 X8133M 7.358-3 XE135 5.888-1 5.1083 3.0083 XE135M 5.91E-1 X8137 2.11 8.1083 1.71E4 X8138 1.93 7.1083 1 ~ 37E4 AR41 TOTALS 8.36 ~ 4.99E4 The Offgas Noble Gas Monitor Alarm Setpoint equation is: Alarm Setpoint ~ 0.8>350,000"2 '8-3>Ei(Ci>CFi)/ff"Ei(Ci)] + Bkg. Where the Alarm Setpoint is in cpm, Ei(Ci>CFi) is 4.99E4 cpm, f is 25CFM and Ei(Ci) is 8 36 yCi/cc. This will yield an alarm setpoint of 1.40E5 cpm above background. Particulates and Zodines are not included ,in this calculation because this is a noble gas monitor. To provide an alarm in the event of failure of the offgas system flow instrumentation, the low flow alarm setpoint will be set at or above 10 scfm, (well below normal. system flow) and the high flow alarm setpoint will be set at or below 120 scfm, which is well above expected steady-state flow rates with a tight condenser.
-36 September 1987
;u; "'.;".uen" s Do 3.ate Caiculatio.".
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) an d
- 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 w5ich 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 ga'ses: 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, i, from, the Stack release averaged over the time period of concern, uCi/se c Ki Is the constant accounting for the whole body dose rate from immersion'n the semi-infinite -cloud for each identified noble gas nuclide, i. . Listed on Table 3-3, mrem/yr per uCi/m3 May 1986
's . the '.".ighes t calculate<i annual average re.'=!" ' =
concentration at or beyond the site boundry for the V~nt. Final Environmental S tatement, HUREG 1085, Table D- 2, 2.0E-6 se c/m3 Qiv Is the release rate of each noble gas nuclide, i, from the Vent release averaged 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 pCi/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 3.11.2.1.a.
3.2.2 Skin Dose Rate Due to Noble Gases This calculation covers TS Section 3. 11.2.l.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'ctivity from the stack to the receptor 'is taken into account by use of the factor (X/Q). Gamma r'adiation exposure from the overhead plume is taken into account by the factor l,lBi. 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: mrem/yr Ej, [ (Li+(X/Q)s + 1.1~Bi)<<Qis + (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
Is the constant accounting zor the air gamma iose from exposure to the overhead plume oz e'levat'-d releases of each identified noble"'as nuclide, i . Listed on Table 3-2, mrad/yr per uCi/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 Environgental Statement, NUREG 1085, Table D-2, 4.5E-8 sec/m Qiv Is the release rate of each noble gas nuclide, i, from the Vent release averaged over the time period "-of concern, >tCi/se c Qis Is the release rate of each noble gas nuclide, i, Stack release averaged over the time period of from'he concern, uCi/se c Example Calculation: Assume an analysis of the Stack and, Vent Effluents indicate that 1.81E4 and 1.26E4 pCi 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 uCi/m3 for 'the Stack and Vent is 4.5E-8 and '2.0E-6 sec/m3 respectively'X/Q) 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. 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 greater 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 Ep [ Zi Pip [WsQis + WvQiv] ) Where: Pip 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 ~>Ci/m . For ground and food pathways, m2~rem/yr per '<i/sec . Zi Is the summation over all nuclides, i Is the summation over all pathways Ws, Wv Are the dispersion parameters for stack and vent release resyectively for each pathway as approriate sec/m 3 or 1/m . See Table 3-22. Qis, Qiv Are the release rates for nuclide i, from the stack and vent respectively pCi/sec. Example Calculation ~ Assume. an analysis of the Stack and Vent -Effluents indicate that 1.84E-1 and .1.26E-1 pCi/sec of I-131 are released from each point respectively. From Ta'ble 3-4. thru 3-6 and, 3-22 the following table be made:
'an ORGAN Pi GROUND Pi INHALATION Pi FOOD or m2~rem/yr mrem/yr m2~rem/y r FACTOR nCi/sec uCi/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
CROUt'D CiHALa"lOÃ FOOD <0<<-' 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 3.11.2.1.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'qual 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. 3.3. 1 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 rel'eases. 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: m<<d ~ [Mi(X/Q)v Qiv + Bi Qisl Where the constants have all been p'reviously defined. Note that since Q is expressed as pCi/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 pCi 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. Ray 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 accountThethefactors dispersion are of releases to the most conservative location. discussed in greater detail later. Beta air dose due to noble gases: Zi"i[(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 1.42Ell and 9.91E10 Si of Xe-133 are released from each point respectively over the last month. This correlates to 1.81E4 and 1.26E4 pCi/sec respectively. From Table 3-3, Ni is 1.05E3 mrad/yr per ~iCi/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 to 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 zp [ Ef Rip [Ws Qis + Wv Qiv] ] t Where: 3; 17E-8 Is the inverse of the number of seconds in a year I Rip Is the factor that takes into account the dose to an i individual organ from nuclide through pathway p. May 1986
fs the summation over all nuclides i.
'-P Is the summation over all pathways p.
Ws, Wv Are the dispersion parameters for the stack and ven t respectively for each pathway as appropriate sec/m or 1/m . See Table 3-22. Qis, Qiv Are the amount of activity of nuclide i released from the stack or vent respectively over the period of 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 gCi of I-131 are released from each point respectively over the last quarter. This correlates 1.84E-l 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-INHALATION Ri-MILK Ri-MEAT Ri-VEGETATION or m2~rem/ r mrem/ r m2-mrem/ r FACTOR uCi sec pCi m3 uCi sec T BODY 1. E E E~ ~~Eh .1 E SKIN 2. 09E7 BONE 4.81E4 6.51E8 8.26E6 1.43E8 LIVER 4.81E4 6.55E8 8.32E6 1.44E8 THYROID 1.62E7 2. 17E11 2.75E9 4.75E10 KIDNEY 7.88E4 1.08E9 1.37E7 2.36E8 LUNG GI-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 KRONE-2 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 2 BODY '111E-2 E- E-2 ~1E-3 MME-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-l 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 "h~s zse =he:".axim<<m quarterl.v dose 'o the child o" ~<,". 11.7 mrem to the thyroid f rom I-131. The calculation woui: repeated for all nuclides and age groups and summed to find th dose to any organ. The dose limit to the maximum exposed 'aximum organ is specified by TS Section 3.11.2.3 and 3.11.2.5. 3.4 Gaseous Effluent Dose Factor Definition and Derivation 3.4.1 Bi and Vi- 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 exponenti'al of'he 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 . 45 May 1986
I,".e hose 0 at a point is proportional to the flux o, energy o t the radiation, air energy absorption coefficient Ua (m-' 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)*Vouch]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-dis~ /MevmCi-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 Tdthe(gm/cc) is assumed to be absorption for 5gm/cc (like a rock) and Ut (cm2/gm) is energy 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 (muscle).'f 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- 4. 7E-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 ODCMo 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, 'ritium and Particulates with Half-lives greater than 8 days. Table 3-4 Ground Plane Pi (m2mrem/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: K' a constant of unit coversion, 10 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) .
DFGi 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 radionuclide, DFGi, are presented in Table 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 DFGi (1-e "it)/>i Example Calculation For the I-131 total body dose rate factor for exposure from the ground: ii ~ 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-mrem/yr per uCi/sec as listed on Table 3-4. It should be noted that only a limited number of nuclide s ar e listed on Table 3-4. These ar e the mos t itIfwill common nuclides encountered in effluents. a nuclide is detected for which a factor is not listed, then be calculated and included in a revision to the ODCM. 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 ehe I-131 skin dose rate factor for exposure from the ground:
~ r
~ 9.98E-7 sec-1 DFGi 3.40E-9 mrem/hr per pCi/m2 These values will yield a Pi factor of 2.98E7 'm2mrem/yr per uCi/sec as listed on Table 3-4. It should be noted that only a limited t
nubmer of nuclides are listed on Table 3-4. These are the mos 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 exposure. (From NUREG 0133, section 5.2.1.1) Pi K'(BR) DFAi (mrem/yr per pCi/m ) Where '. K' constant of unit conversion, 106 pCi/)Ci. BR ~ the breathing rate of the infant age group, in m3/yr. DFQ 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 rate is taken as 1400 m /yr from Table E-5 of Regulatory Guide 1.109. The inhalation dose factors for the infant, DFAi are presented in Table E-'10 o f Regulatory Guide 1.109, in units o f mre m/pCi ~ Resolution of the units yeilds:
'i (inhalation) 1.4 x 10 DFAi.
Example Calculation: Por 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. If a nuclide is detected for. which a factor is not listed, then it will be calculated and included in a revision to the ODCM. Table 3-6, Pood (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; (Prom NUREG 0133, s ection 5.2. 1.3)
~ C'r q(u,
~s' )
F DFL< (~ S rj (~ .erm/sr eer <<Ci/sec) p 1 w Where . 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 Pm the stable element transfer coefficients, in days/liter. fraction of deposited activity retained on cow's feed grass.
\
DPLi . the maximum organ ingestion dose factor for the ith radionuclide, in mrem/pCi.
-1 Xi ~ the decay constant for the ith radionuclide, in sec the decay constant for removal plant surfaces by weathering,
'f activity on leaf and 5.73 x 10 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 , and Y are provided in Regulatory Guide 1.109, Tables E-3, E-E, and E-E5, as 50 kg/day, 330 liters/day and 0.7 kg/m , respectively. The value tf is provided in Regulatory Guide 1.109, Table E-15, as 2 days (1.73 x 105 seconds). 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:
;tcod j ~ 2.4xlOse '~- lsd., $e 1 t'] (m'ran/yr per ..'. f/sec) for all radionuclides, except tritium.
The concentration of tritium in milk is based on its airborne concentration rather than the deposition rate.
. p~ ~ K'K'"F g>U OfL< $ 0.75(0.5/H)) (mree/yr per >C1/as)
Where: K'" a 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 2 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 10 mrem/yr per uCi/m for tritium, only Example Calculation'. For I-131 thyroid does rate factor for exposure from cow milk ingestion. May 1986
r ~ 1.0 unitless for Iodines Fm 6E-3 days/liter DFLi 1.39E-2. mre m/pCi ii = 9.98E-7 sec-1 iw 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 it will be calculated and included in a revision to the ODCM. 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 (DPAi)a (mrem/yr per uCi/m ) Where: K' constant o f unit conversion, . 10 pCi/ pCi ~ (BR)a th~ breathing rate of the receptor of age group (a),in m /yr. (DPQ)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 selection 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 Breathin Rate (m ~/ r)
'nfant 140 0 C}lild 3700 Teen 8000 Adult 8000 Inhalation dose factors (DPAi)a for the various age groups are given in Tables E-7 throught E-10 of Regulatory Guide 1.109.
Example Calculation: r Por the I-131 infant thyroid dose factor for exposure from inhalation: DPAi 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 Xf a nuclide is detected it
~
f or 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)DFGi f (1-e "i ) />i ] (m mrem/yr per uCi/sec) Where: K' constant of unit conversion, 10 pCi/)lCi. Ka constant of unit conversion, 8760 hr/year. 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 radionuclid (mre'm/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: M ~ 9.98E-7 sec-1 DFGi 2.80E-9 mrem/hr per pCi/m2 These values will yield a Ri factor of 172E7 m2~rem/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 nuclide s 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'i (m2~rem/yr per uCi/sec) also takes'nto 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:
)i' 9.98E-7 sec-1 DFGi 3.40E-9 mrem/hr per pCi/m2; These values will yield a Ri factor of 2.09E7 m2mrem/yr per uCi/sec as listed on Table 3-11. It should be noted that only a limited number o f nuclides are listed on Table 3-11. These are the mos t common nuclides encountered in effluents.
for which a factor is not listed, then itIfwill a nuclide is detected be calculated an included in a revision to the ODCM. TABLES 3-12 to 3-15 Ri VALUES COW MILK Ri (m2~rem/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). 5 (4 2
~~yr per gC)/sec)
Where. K' a constant of unit conversion, 106 pCi/pCi ~ Qp the cow's consumption rate, in kg/day (wet weight). Uap
~ the receptor' milk consumption rate, in 1 iters/yr .
Y P the agricultural productivity by unit area of pasture feed grass, in kg/m 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; Xi ~ the decay constant for the ith radionuclide, in sec -1 May 1986
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). tf = the transport time from pasture receptor, in sec. to cow, to milk, to the transport time from pasture, to harvest, to cow, to milk, to receptor, in sec. fp = fraction of the year that the cow is on pasture (dimensionless). fs = 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 value of fs will be consi,dered unity. TCN i fp will be considered to be 0.5 for a May to October grazing season. 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 Table r (dimensionless) 1.0 for radioiodine E-15
- 0. 2 for particulates E-15 Fm (days/liter) Each stable element E-1 Ua (liters/yr) Infant 330 E-5 Child '30 E-5 Teen 400 E-5 Adult 310 E-5
( FLi)a (mrem/pCi) Each radionuclide E-11 to" E-14 Yp (kg/m2) 0.7 E-15 Ys (kg/m2) 2.0 E-15 tf (seconds) 1.73 x 10 7.78 x 10 (2 days) (90 days) E-15 E-15 th (seconds) 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]: Ri = K'K"FmQFUa DFLi[0.75(0.5/H)) (mrem/yr per vCi/m ) May 1987
Nhere: K" a constant of unit conversion, 10 gm/kgb H absolute humidity of the atmosphere, in gm/m 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 ii iw 9.98E-7 sec -1 5.73E-7 sec -1 tf 1.73E+5 sec. These values will yield a factor of 5.26Ell and 6.31E11 mrem/yr per uCi/sec respectively for cow and goat milk. However, the actual dose to the infant thyroid is also dependant'n 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 is relatively so much smaller than the slightly larger Ri goat'eposition 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 5.26Ell mrem/yr per uCi/scca 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 CON 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) S Q<tvq] ~ r.','<(I t)~(u, )
)(orL,), ~ p s
~a' 5
"<'r (e 2 ~~/~", pet'CI/sec)
-'55-. May 1986
Where: Ff the stable element transfer coefficients, in days/kg. Uap the receptor ' meat consumption. rate for age ( a),'sec. in kg/yr. tf ~ the transport t ime from pasture to receptor, in th ~ the transport time from crop field to receptor, in sec. Tabulated below are the appropriate'arameter 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 "ap (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/m~) 0.7 E-15 Ys (kg/m2) 2.0 E-15 tf (seconds). 1.73 x 106 (20 days) 7 78 x 106 (90 days)
E-15 E-15 th (seconds) E-3 Qp (kg/day) 50 The concentration of tritium in meat is based on the airborne concentration rather than the deposition. Therefore, the Ri is based on [x/Q]: R)
"'"'" gQ"~p ) ~ fo (o. /H)j (area/yr per gCi/Q) 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 mre m/pCi ~ These values will yield a Ri factor of 2.75E9 m2mrem/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 most common nuclides encountered in effluents. a nuclide is 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 (m2mrem/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 th'e 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: ik I h ia [ ULf + USf e I v i (0FI ) at e I(m 2 mrem/yr per 11Ci/sac)
<<here:
6 K'. a constant of unit conversion, 10 pCI/11CI. V Ua the consumption rate of fresh leafy vegetation by the receptor in age group (a), in kg/yr. the consumption rate of stored vegetation by the receptor in age group (a). in kg/yr. r~ the fraction of the annual intake of fresh leafy vegetation grown>locally . fg ~ the fraction of the annual intake of stored vegetatian grown locally. the average time between harvest of -leafy vegetation and its cans1snption, in seconds.'he average time between harvest of stored vegetation and its consumption, in seconds. Yv the vegetati on area 1 dens I ty, I n kg/m> ~ and all other factors are defined in this manual ~ Tabulated below are .the appropriate parameter values and their reference to Regulatory Guide 1.109. Parameter Table r (dimensianless) 1.0 for radioiodines E-l 0.2 for particulates E-l 0FI.i (mrem/pCi ) Each radionuclide E-ll 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
.i'L (dimensianless) site specific (default ~ 1.0) f (dimensianless) site specific (default ~ '0.76) (see QGIA:IXgpage 2B) t< (seconds) 8.6 X 10 (1 day) E-15 th (seconds) 5.18 X 10 (60 days) E-15 (kg/m2) 2.0 E-'I 5 Yv 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]: a<(da] ~ x'x' L 5 [0.75(0.5/H)3 (mrenlyr per L a (OFL<)~ ~C</m ). 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'. r 1.0 unitless for Iodines DFLi ~ 5.72E-3 mrem.pCi. These values will yield a Ri factors o f 4.75E10 m2~rem/y r pe r
<<Ci/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 it will be calculated and included in a revision to the ODCM.
3'4.5 X/Q and Wv - Dispersion Parameters for Dose Rate, Tab3.e 3-22 The dispersion parameters for the whole body and skin dose rate 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 is higher than any ad]scent building it is not more than 2.5 times the height. 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 It not all exist at the same location. is conservative to assume that a single individual would actually be at each of the receptor locations. May 1986 e
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 7B, as
.. noted in Section 3.4.5. These were calculated using the methodology of Regulatory Guide l.'111 and NUREG 0324. The Stack was modeled as an elevated release point because its height is mor'e than 2.5 times than any adjacent building. The Vent was modeled as a combined elevated/ground level release because even though it is higher than any adjacent building it is not more than 2.5 Average meterology over the appropriate time period was used.
times the height 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. I-133 Estimation The Stack and Vent Effluent Monitor at Nine Mile Point-Unit 2 are isotopic monitors. They are designed to automatically collect on'ine iodine samples'n 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-131 concentration. The ratio will be determined. at least quarterly by analysis af short duration samples .
3..6 Use of Concurrent Meteorological Data vs.'istorical 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'o 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 e)ector 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 particulate release, t'o be OPERABLE are:
- 1) HEPA Filter Radwaste Decon Area
- 2) HEPA Filter Radwaste Equipment Area
- 3) HEPA Filter Radwaste General of these filters is not OPERABLE, iodine and particulate Area'henever one 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
TABI E 3-1 OFFGAS PRETREATMENT" DETECTOR RESPONS E NUCLIDE NET CPM/ uCi/cc Kr 85 4.30E+3 Kr 85m 4.80E+3 Kr 87 8.00E+3 Kr 88 7.60E+3 Xe 133 l. 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 -: uCi/sec)
Kr 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 l. 51E-2 Xe 131m 6.90E-5 6.55E-5 Xe 133 6.12E-4 5. 9 3E-4 Xe 133m 3.62E>>4 . 3.44E-4 Xe 135'e 4.31E-3 4. 09E-3'.12E-3 135m 6.55E-3 Xe 137 3. 07E-3 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 DOS E FACTORS Nuclide ~K ((y-Body)** L (B-Skin)** M)(y-Aic)e** Ni(B-Aic)e** 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 1.72E1 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'.48E3 Xe 133m 2.51E2 9.94E2 3. 27E2 Xe.133 2.94E2 3.06E2 3.53E2 1.05E3. Xe 135m 3.12E3 7.11E2 3.36E3 ~ 7.39E2 Xe 135 1.81E3 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 ~
3 "**mrad/yr per t<Ci/m . May 1986
TABLE 3-4 Pi VALUES GROUND PLANE**
>>Ci/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 9.99E8
- 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 accord ance with NUREG 0133, Section 5.2.1.2. May 1986
TABLE 3-5 P i VALUES INHALATION** m~reml r 3 uci/m NUCLIDE BONE I IVER T. BODY THYROID 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 8.95El 5.75E1 1.32El 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 'SE4 6.47E5 5.14E4 Sr 89 3.98E5'r 1.14E4 2.03E6 6.40E4 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 1.33E3 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 CG 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 3-6i VALUES FOOD (Cow Milk~++* m 2 - mrem/yr, . uCi/sec NUCLIDE BONE LIVER T. BODY THYROID 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 S.SOE6 1.46E7 Fe 59 2.28E8 3.99ES 1.57E8 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 1. 64E10 S r 89 l. 28E10 3.6>ES 2.63ES ~
Sr 90 1.24Ell 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 2.12ES 4.13E7 3.17ES . 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 6.92ElO 6.99E9 1. 78El0 7. 31E9 1. 88ES Cs 137 5.24E10 6.13E10 4.35E9 1.65E10 6.67E9 1.92E8 Ba 140 2.45ES 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 1.33E5 3.92E5 1.36E8
~mrem/yr per pCi/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 INHALATION INFANT<* m~rem/ e 3 uCi/m NU GLIDE ~ 30NE LIVER I ~ BODY THYROID KIDNEY LUNG 6.47E2'I-LLI H 3 6.47E2 6.47E2 6.47E2 6.47E2 '.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 1.11E4 Co 60 8.02E3 1.18E4 4.51E6 3.19E4 Zn 65 1.93E4 6.26E4 3.11E4 3 '5E4 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 6.4 3E3 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 '.'33E3 Ba 140 5.60E4 5.60E1 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 '6E4 Ce 144 3.19E6 1.21E6 1.76ES 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
. i VALUES INHALATION - CHILD m~rem/ r 3
><Ci/m NUCLIDE BONE LIVER T. BODY THYROID KIDNEY I DNG G I-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 m e 1.00E4 1.58E6 2.29E4 Fe 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
'r 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 4.26E1 3D92E2 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. 11E1 1.74E6 1.02E5
+La 140 6.44E2 2.-25E2 7.55E1 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 pCi/m NUCLIDE BONE LIVER Y. BODY: THYROID KIDNEY LUNG G I-LLI; H 3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3 1.27E3'r C 14 2.60E4 , 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3 4.87E3" 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 1.78E5 '.53E6 Co 58 2.07E3 2.78E3 1 e'34E6 9.52E4
'Co 60 1.51E4 1.98E4 2.59E5 '.72E6 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'"'.69E5 Mo 99 1.69E2 3.22E1 4 11E2 1.54E5 I 131 3.54E4 4.91E4 2.64E4 1.46E7 8.40E4 6.49E3'.03E4.
I 133 1.22E4 2.05E4 6.22E3 2.92E6 3.59E4 Cs 134 5.02E5 1.13E6 5.49E5 3.75E5 '.46E5 9.76E3Y Cs 137 6.70E5 8.48E5 3.11E5 3.04E5 1.21E5 8.48E3'.28E1 Ba 140 5.47E4 . 6.70E1 3.52E3 - 2A3E6 . 2.29E5
- La 140 4.79E2 2.36E2 6.26El 2. 14E5'.87E5 Ce 141 2.84E4 1.90E4 2.17E3 8.88E3 '.14E5 1.26E5 Ce 144 4'89E6 2.02E6 2'62E5 1.21E6 1.34E7 '.64E5
- Daughter Decay Product. Activity level and effective half life =assumed to equal parent nuclide.
"69- May 1986
TABL E 3-10 INHALATION ADULT
.i VALUES R
m~rem/ r 3 gCi/m NUCLIDE BONE LIVER Y. BODY YHYROID KIDNEY IUNG 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. 9'6E4 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 '0E6 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 2.30El, 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 '7E5 9.76E4 1.04E4 Cs 137 4.78E5 6.21E5 4.28E5 2.22E5 7.52E4 8.40E3 Ba 140 3.90E4 4.90El 2.57E3 1'67El- 1.27E6 2.18E5
- La 140 3.44E2 .'.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 2 m mrem/yr -'. uCi/sec 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 95 2.45E8 2.85E8
*Nb 95 '.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.
"71- May 1986
TABLE 3-12 R i VALUES 2 COW MILK INFANT m ~rem/yr -'. pCi/sec NUCLIDE BONE LIVEN T. BODY THYHOID 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 1.22E8 2.13E8 8.38E7 6.29E7 1.02E8 Co 58 1.39E7 3.46E7 3.46E7 Co 60 5.90E7 1.39E8 1.40E8 Zn 65 3.53E9, 1.21E10 5.58E9 5.87E9 1.02E10 Sr 89 6.93E9 1.99E8 1.42E8 Sr 90 8.19E10 2 09E10 1.02E9 Zr 95 3.85E3 9.39E2 6.66E2 1.01E3 4.68E5
- Nb 95 3.93ES 1.62E5 9.35E4 1 16E5 1.37E8 Mo 99 1.04E8 2.03E7 1.55E8 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 I41 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 CHIT D 2 m ~rem/yr -.'iCi/sec NUCLIDE BONE LIVER Y. BODY IHYROID 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 2.93E4 7.99E3 5.34E4 2.80E6 1.13E7'.27E4 Mn 54 1.35E7 3.59E6 3.78E6 Pe 59 6.52E7 1.06E8 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 1.41E8 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 1.52E8 Mo 99 4.07E7 1.01E7 8.69E7 3.37E7 I 131 6.51E8 6.55E8 3.72E8 2.17Ell 1.08E9 5.83E7 I 133 8.58E6 1.06E7 4.01E6 '.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 '1E3 7.15E6 Ce 144 1.04E6 3.26E5 5.55E4 1.80E5 8.49E7 "mrem/yr per aCi/m ~
- Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-14 R VALUES COW MILK TEEN 2 m harem/yr . pCi/sec NUCLIDE BONE LIVER T. BODY THYROID 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 . 1.75E8 Sr 90 4.4 5E10 1.1QE10 1.25E9 'r 95 9.34E2 2.95E2 2.03E2 4.33E2 6.80E5 +"Nb 95 9.32E4 5.17E4 2.85E4 '.01E4 2.21E8 Mo 99 2.24E7 4.27E6 5.12E7 4.01E7 I 131 2.68E8 3.76E8 2.02ES 1.10E11 6.47E8 7.44E7 I 133 3.53E6'.99E6 1.83E6 . 8.36E8 1 .05E7 4.53E6 Cs 134 6.49E9 1.53E10 7.08E9 4.85E9 1.85E9 1.90E8 Cs 137 9 02E9 1.20E10 4.18E9 4.08E9 1.59E9 1.71E8 Ba 140 '.43E7 2.98E4 1.57E6 1.01E4 2.00E4 3.75E7
- +La 140 3.73E1 1.83E1 4.87EO 1 .05E6 Ce, 141'.67E3 3.12E3 '.58E2 1.47E3 8,91E6 Ce 144 4.22E5 '.74E5 '.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 R VALUES COW MILK - ADULT 2 m ~rem/yr -- UCi/sec NUCLIDE BONE LIVER T. BODY THYROID 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 Pe 59 1.61E7 3.79E7 1.45E7 1.06E7 1.26ES Co 58 2.70E6 6.05E6 5.47E7 Co 60 1.10E7 2.42E7 2.06E8 Zn 65 8.71E8 2.77E9 1.25E9 1.85E9 1. 75E9 Sr 89 7.99ES . 2.29E7 1.28E8 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 1.84ES Mo 99 1.24E7 2.36E6 2.81E7 2.87E7 I 131 1.48ES 2.12ES 1.21ES 6.94E10 3.63ES 5 '8E7 I 133 1.93E6 3.36E6 1.02E6 4.94ES 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 . 1.32E8 Ba 140 1.35E7 1.69E4 8.83E5 5.75E3 9.69E3 2.77E7
"*La 140 2.07E1 1.05E1 2.76EO 7.67E5
'e 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/m3.
- 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/sec NUCLIDE BONE LIVER Y. BODY IHYROID KIDNEY LUNG GI-LLI "H 3 2.34E2 2.34E2 2.34E2 . 2.34E2 2.34E2 2.34E2 "C 14 '.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 '.37E6 1.44E6 4.32E6 Fe 59 2.04E8 3.30E8 '.65E8 9.58E7 3.44E8 Co 58 9.41E6 2. 88E7 5.49E7 Co 60 '.64E7 1.37E8 2.57E8 Zn 65 '2.38E8 6.35E8 3.95E8 4.00E8 1.12E8 Sr 89 '2.65E8 7.57E6 1.03E7 Sr 90 7. 01E9 1.78E9 9.44E7 Zr 95 1.51E6 '.32E5 2.95E5 4.75E5 .3.46E8 .
- 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 'l92E-1 1.43E-1 Cs 134 6.09E8 1.00E9 2.11E8 3 ~ 10E8 1.11E8 5 ~ 39E6-Cs 137 8.99E8 8.60E8 1.27E8 2.80E8 1 .01E8 5.39E6 Ba 140 2.20E7 1.93E4 1.28E6 6 '7E3 1.15E4 1.11E7
++La 140 1.67E2 '.84E1 1.97E1 1.63E6 Ce 141 1.17E4 5.82E3 '.64E2 2.55E3 7.26E6 Ce 144" 1.48E6 4.65E5 7.91E4 2.57E5 1.21E8
+mrem/yr per pCi/m . <<"Daughter Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-17 R VALUES COW- MEAT TEE N 2 m -mrem/y -'. iiCi/sec NUCLIDE BONE LIVER T. BODY THYROID 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 '.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 1.15E8 2.69E8 1 .04E8 8.47E7 6.36E8 Co Co 60'.05E6 58 3.90E7 1.86E7 8.80E7 1.11E8 5.09E8 Zn 65 1.59E8 5.52E8 2.57E8 3.53E8 2.34E8 Sr 89 1.40E8 4.01E6 1.67E7 Sr 90 5.42E9 3,.34E9 1.52E8 Zr 95 '.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-1 8.00E-2 "3.66E1 . 4.60E-1 1.99E-1 Cs 134 3.46E8 - 8.13E8 . 3.77E8 2.58E8 9.87E7 1.01E7 ~ Cs 137 4.88E8 6.49E8 2.26E8 2.21E8 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
'.14E3
'e 141 6.19E3 4.75E2 1.95E3 1.18E7 Ce 144 7.87E5 . 3.26E5 4.23E4-=- 1.94E5 1.98E8-
- mrem/yr p'er pCi/m3.
**Daughter'Decay Product. Activity level and effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-18 Ri VALUES - COW MEAT - ADULT 2 m harem/yr ='Ci/sec NUCLIDE BONE LIVER Y. BODY 'IHYROID KIDNEY LUNG GI-LLI
- H 3 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2 3.25E2
- C 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 1.44E8 3.39E8 1.30E8 9.46E7 1.13E9 Co 58 1.04E7 2.34E7 2.12E8 Co 60 5.03E7 1.11E8 9.45E8 Zn 65 ~ 2.26E8 7.19E8 3.25E8 4. 81E8 4.53E8 Sr 89 1.66E8 4.76E6 2;66E7 Sr 90 8.3,8E9 2.06E9 2.42E8 e
Zr 95. 1.06E6 3.40E5 2.30E5 . 5.34E5 1.08E9
- Nb 95 1.79E6 9.94E5 5.35E5 9.83E5 6 '4E9 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-1 3.22E-l 9.81E-2 4.73E1 . 5.61E-1 2.89E-1 Cs 134 4.35E8, 1.03E9 8.45E8 3.35E8 1.11E8 l. 81E7 Cs 137 5.88E8 8.04E8 5.26E8 2.73E8 9.07E7 1.56E7 Ba 140 1.44E7 1.81E4 9.44E5 6 ~ 15E3 1. 04E4 2. 97E7
- "La 140 1.11E2 5.59E1 1.48E1 4.10E6 Ce 141 7.38E3 4.99E3 5.66E2 2 '2E3 1.91E7 Ce 144 9.33E5 3.90E5 5.01E4, 2.31E5 3. 16E8
- mrem/yr per iiCi/m
**Dauahter Decay Product.
equal parent nuclide. Activity level
.-?8- May 1986 and effective half life assumed to t
TABLE 3-19 R VALUES VEGETATION CHILD 2-m ~rem/yr -'. pCi/sec NUCLIDE'ONE LIUER Y. BODY YHYROID KIDNEY LUNG GI-LLI
*H 3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 4.01E3 "C 14 3.50E6 7.01E5 7.01ES 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 5.58ES Fe 59 3.97ES 6.42ES 3.20E8 1.86E8 6.69E8 Co 58 6.45E7 1.97ES 3.76ES Co 60 3.78E8 1.12E9 2.10E9
'n 65 8.12ES 2.16E9 1.35E9 1.36E9 3.80ES
'r 89 3.59E10 1.03E9 1.39E9 Sr,90 1.24E12 3.15Ell 1.67E10 Zr 95 3.86E6 8. 50E5 7.56E5 1.22E6 8.86E8
**Nb 95 7 50E5 2.92E5 2.09E5 2.74E5 '.40E8.
Mo 99 7.70E6 1.91E6 1.65E7 6.37E6 I 131 1.43ES 1.44E8 8.16E7 4.75E10 2.36ES 1.28E7 I 133 3.52E6 4.35E6 '1.65E6 8.08E8 .7.25E6. 1.75E6 Cs 134 1.60E10 2.63E10 5 5SE9 8.15E9* 2.93E9. 1.42ES Cs 137 2.39E10 2.29E10 3'38E9 7.46E9 2.68E9 1.43ES Ba 140 2.77E8 2.43ES 1.62E7 7. 90E4 = 1.45E5 1.40E8
**La 140 3.37E4 1.18E4 3.97E3 3.ZSES Ce 141 6.56E5 3.27E5 4.85E4 1.43E5 - ~ 4.08ES Ce 144 1.27ES 3.98E7 6.78E6 2.21E7 "
1,04 E10 =
*mrem/yr per. nCi/m3. <<*Daughter Decay Product. Activity level and 'effective half life assumed to equal parent nuclide.
May 1986
TABLE 3-20 R i VALUES VEGETATION TEE
-2 N
m ~rem/yr -'. uCi/sec NUCLIDE BONE LIVER Y. BODY IHYROID 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. 54E8 9. 01E7 1.36E8 9.32ES Fe 59 1.79ES 4.18E8 1.61E8 1.32ES 9.89ES
,Co 58 4.37E7 1.01E8 6.02E8 Co 60 2.49E8 . 5.60E8 3.24E9 Zn 65 4.24ES 1.47E9 6.86E8 9.41E8 6.23E8 Sr 89 1.51E10 4.33E8 1.80E9 Sr 90 7.51E11 1.85Ell 2.11E10-Zr 95 1.72E6 5.44E5 3.7.4E5 7.99E5 1.26E9
- Nb 95 3.44E5 1.91E5 . 1.05E5 1.85E5 8.16E8 Mo 99 5.64E6 1.08E6 1.29E7 1.01E7 I 131 7.68E7 1.07E8 5.78E7 3.14E10 1.85E8 2.13E7 I 133 '.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 1.92E8 Ba 140 1.38E8 1.69E5 8.91E6 5.74E4 1.14E5 2 '3ES
- 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 4
~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 harem/yr .'Ci/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 1.79ES . 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. 58ES Pe 59 1.26E8 2.96E8 1.13E8 8.27E7 1.02E9 Co 58 3.08E7 6.90E7 6.24E8 Co 60 1.67E8 3.69E8 3. 14 E9 Zn 65 3. 17E8 1.01E9 4.56E8 6.75E8 6.36ES Sr 89 9.96E9 2.86E8 1.60E9 Sr 90 6.05Ell 1.48Ell 1.75E10 Zr 95 1.18E6 3.77ES 2.55E5 . .5.92E5 4'
,1.20E9
- +Nb 95 2.41E5 1.34E5 7.20E4 1.32E5 8.13E8 Mo 99 6.14E6 1.17E6 1.39E7 1.42E7 I 131 8. 07E7 1. 15E8 6. 61E7 3. 78E10 1. 98ES 3.05E7 I 133 2.08E6 3.61E6 1. 10E6 5.31E8 6.30E6 3.25E6 Cs 134 4.67E9 1.11E10 9.08E9 3.59E9 1;19E9 1.94E8 Cs 137 6.36E9 8.70E9 5.70E9 2.95E9 9.81ES 1.68E8 Ba 140 1.29E8 1.61E5 8.42E6 5.49E4 9.25E4 2.65E8
- ~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 pCi/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 W VALUES VENT DIRECTION DISTANCE (m) X/ (sec/m ) U~l(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 TACK Site Boundary*"*.. 1$ 600. 4.50Z-8 6.00E-9 Inhalation and Ground E (109') 1,700 8.48E-9 j. 34E-9 Plane Cow Milk ESE (135') 4,200 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.
"*>> X/Q and D/Q from NMP-2 FES$ NUREG-1085$ May 1985.
May 1986
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4.0 UR.MXIUif 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 lightmater-cooled nuclear power plant using uranium fuel, and reprocessing of spent uranium fuel, to the extent that these directly support the production of electrical power operations, for public use utilizing nuclear energy, but excludes mining operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered nonmranium special nuclear and by-produc t 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 such that the dose Report to the NRC and limit subsequent releases 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 gaseou's 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
'imits have occurred,'then a variance from the NRC to permit such releases will be requested and if possible, action. will be taken to reduce subsequent releases.
The report to the NRC shall contain: Identification of all uranium fuel cycle facilities o r 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- The 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 t'o determine a pro)ected direct dose component. 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'd) Por the purpose of calculating doses, the results of the Environmental Mc(nitoring Program may be included to provide more refined 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 Prom 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. Because of the sample medium type and the half-lives of the -radionuclides historically
.observed, the decay corrected portions of the equations are deleted.
This, does not reduce che conservatism of the calculated doses but increades the simplicity from an evaluation point of view. The dose from fish sample media is calculated as: (1) Rwb Ei [Cif x u x 1000 x Diwb x 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-ll) i for the whole body of an The fractional portion of the year over which the dose is applicable. (2) Rl Ki [Cif x ux1000x Dilx f] August 1986
Where: Rl = The total dose to the liver of an adult (maximum exposed organ) in mrem per year Cif The concentration of radionuclide i in fish samples in pCi/gram p 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-11) i for the liver of an
/
The fractional portion of the year over which the dose is applicable. The dose from shoreline sediment sample media is calculated as: Rwb Ei [Cis u x 40,000 x 0.3 x Diwb x and Rsk Ei [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 in concentration of radionuclide i in shoreline sediment pCi/gram The usage factor. This is assumed as 67 hours per year by a teenager 40,000 The product o f f the assumed density o shoreline sediment (40 kilogram per square meter to a depth of 2.5 cm) times f the number o 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 i 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 Effluent s 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 radwast'e 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 results
.92- . May 1986
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' ground plane dose (deposition). are not applicable. 'n Other pathways, such as the ingestion pathway, addition, pathways associated with water related recreational activities are not applicable here. These include 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/Q 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 3 /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] August 1986
4.4 (Cont'd) where R the dose for the period in question to the lung (j) for all radionuclides (i) for the adult age group (a) in mrem per time period . Ci The concentration in the stack release of radionuclide i inaverage pCi/m for the period in question F = Average effluent flowrate in m /sec . X/Q The plume dispersion parameters for 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) . DFAija the inhalation dose factor for radionuclide i, lung j, 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 will readings't least two envi'ronmental TLD locations 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 adjusted 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 Stations 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 I/17; 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 three units have been calculated for a 5 year period, 1978 through 1982. 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. Specific sample media for which EPA Cross Check Program samples are available include the following: gross beta in air particulate filters 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 August 1986
S 5.3 Capabili'ties for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used users for environmental measurements required by 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'.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 shal'1 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 standax'd deviations of the responses shall be less than 3.0% ~ 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 p'eriod equal to half the same field cycle in an area where the exposure'ate 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 25% 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'o accomplish this, the TLD shall be rotated through at least two perpendicular planes. The response averaged over all directions shall not diffex'rom the response obtained in the standard calibration position by more than 10%. A total of at least 4 TLDs shall be evaluated. May. 1986
Light dependence shall be determined by placing TLDs in the field fo 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 lOX. 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 ~a p 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'E 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) 6 North Shoreline Area (75) 0.1 mi 8 O'. Direct Radiation (TLD) North Shoreline Area (76) 0.1 mi 8 25'.2 Direct Radiation (TLD) 8 North Shoreline Area (77) mi 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.Omit 90 E Direct Radiation (TLD) ll Rt. 29 (79) 1.1 mi 8 115'SE Direct Radiation (TLD) 12 Rt. 29 (80) '.4 mi 8 133'E Direct Radiation (TLD) 13 Miner Road (81) 1.6 mi 8 159 SSE Direct Radiation (TLD) 14 Miner Road (82) 1.6 mi 8 181' Direct'adiation (TLD) 15 Lakeview Road (83) 1.2 mi 8 200'SW Direct Radiation (TLD) 16 Lakeview Road (84) 1.l mi 8 225'W Direct Radiation. (TLD) '7 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 Radiologic'al Environmental Monitoring Program Sampling. Locations Table 5.1 (Continued) Type of
- Map Sample Location Collection Site (Env. Program No.) Location Direct Radiation (TLD) 19 North Shoreline (85) 0.2 mi 8 294'NW Direct Radiation (TLD) 20 North Shoreline (86) 0.1 mi 8 315'W Direct Radiation (TLD) 21 North Shoreline (87) O.l mi 8 341'NW Direct Radiation (TLD) 22 Hickory Grove Road (88) 4.5 mi 8 97',E Direct Radiation (TLD) 23 Leavitt Road (89) 4.1 mi 9 111'SE Direct Radiation (TLD) 24 Rt. 104 (90) 4.2 mi 8 135 SE 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.'3 (93) 4.4 m'i 8 205'SW Direct Radiation (TLD) 28 Co. Rt. 1 (94) 4.7 mi 8 223'W Direct Radiation (TLD) 29 Lake Shoreline (95) 4. 1 mi 8 237'SW t 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'W 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 SE Direct Radiation (TLD) 35 New Haven, NY (56) 5.3 mi 8 123 ESE 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 *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 Fish 42 NMP Site Discharge Area '.6 and/or mi 8 55'E Fish 43 Oswego Harbor Area 6.2 mi 8 235'W Milk 44 Milk Location 850 9,3mi 893 E Milk Milk Location 87 5.5 mi 8 107'SE Milk 46 Milk Location 816 5.9 mi 8 190' Mi'lk 47 Milk Location /140 15.0 mi 8 223'W Food Product 48 Produce Location 86** 1.9 mi 8 143'E (Bergenstock) Food Product 49 Produce Location 81** 1.8 mi 8 96' (J. Parkhurst) Food Product 50 Produce Incation 82*" 1.9 mi 8 101' (Fox) Food Product 51 Produce Location 85** 1.5 mi 8 114'SE (C. ST Parkhurst) Food Product 52 Produce Location 83"" 2.3 mi 8 122'SE (T. Parkhurst) Food Product 53 Produce Location 84"* 2.2 mi 8 123'SE (C. Iawton) Food Product 54 Produce Loc'ation /l7** 15.0 mi 8 223'W (Mc Millen) Food Product 55 Produce Location 88** 126 mi 8 225 SW (Denman)
- Map See Figures 5.1-1 and 5.1-2
"* Food Product samples need not necessarily be collected from all listed locations'ollected samples will be of the highest calculated site average D/Q. 100- May 1986
6.0 DISCUSSION OF TECHNICAL SPECIFICATION REFERENCES Section 3. 12.1 of 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 Representative composite sample aliquots are obtained from 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-intervals 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 cyuse ground water to flow in a northerly direction to Lake Ontario ~ The results of such hydraulic gradient and recharge. prop'erty 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 coriclude 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.
-IOI- May 1986
6.3 Table 3.12-1, footnote Currently, there are no drinking water sources (from Lake Ontario) that can be sign'ificantly affected by the site under normal operating conditions. The closest drinking water source i,s near the City of Oswego. This source is located in an "up-current" 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 source. 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. These 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 'ound 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 ar'e 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 basis. In-frequent use of lake water is not considered to have a significant effect on food products. Therefore, 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- August 1986
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NOTES TO FIGl;RE 5.1.3-1 (a) NMP1 Stack (height is 350') (b) NMP2 Stack (height is 430') (c) JAFNPP Stack (height is 385')
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(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'bove 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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ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1987) NINE MILE POINT NUCLEAR STATION kP2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY JANUARY DECEMBER Doses to members of the public (as defined by the Technical Specifications) from the operation of the NMP2 facility as a result of activity inside the site boundary is controlled by activities at the Energy Information Center. This facility is open to the public and offers educational information, summer picniking activities and fishing. Any possible doses received by a member of the public by utilizing the private road that transverses the east and west site boundaries are not considered here since it takes a matter of minutes to travel the distance. The activity at, the Energy Information Center that is used for the dose analysis is fishing because it is the most time consuming. Although there is no specific survey information available, many of the same individuals have been observed to return again and again because of the access to salmonid and lake trout populations. Dose pathways considered for this activity include direct radiation, inhalation and ground dose (shoreline sediment or soil). Other pathways, such as ingestion pathways are not considered because they are either not applicable or are insignificant. Only releases from the NMP2 stack and vent were evaluated for the inhalation pathway. The direct radiation pathway is evaluated in accordance with the methodology found in the Offsite Dose Calculation Manual (ODCM). This pathway considers radiation from any possible overhead plumes, and direct radiation from plume submersion. The direct radiation pathway is evaluated by the use of high sensitivity environmental TLDs. Since any significant fishing activity near the Energy Information Center occurs between April through December, environmental TLD data for the approximate period of April 1 December 31, 1987 was considered'ata from two environmental TLDs from the approximate area where the fishing occurs were compared to three control environmental TLD locations for the same time period. The average fishing area TLD dose was 8.58E-03 mRem per hour for the period. The average control TLD dose was 7.46 E-03 mRem per hour for the period (approximate second, third and fourth calendar quarters of the year). The average increase in dose as a result of fishing in this area is 3.49 E-01 mRem from direct radiation for the period in question. The majority of the dose from this pathway is from the NMP1 facility. A small portion may be due to the NMP2 facility. The inhalation dose pathway is evaluated by utilizing the inhalation equation in the Offsite Dose Calculation Manual, as adapted from Regulatory Guide 1.109. The equation basically gives a total inhalation dose in mRem for the time period in question (April December). The total dose equals the sum, for all applicable radionuclides, of the NMP2 stack and vent release concentration, times the average NMP2 stack and vent flowrate, times the applicable five year average calculated X/Q, times the inhalation dose factors from Regulatory Guide 1.109, Table E-7, times the Regulatory Guide 1.109 annual air intake, times the fractional portion of the year in question. In order to be slightly conservat,ive, no radiological decay is assumed.
-22
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ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1987) NINE MILE POINT NUCLEAR STATION //2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY (Continued) The 1987 calculation utilized the following information: NMP2 stack: Unit 2 average stack flowrate = 38.23 m3/sec X/g value = 9.6E-07 (annual historical average) Inhalation dose factor = Table E-7 of Regulatory Guide 1.109 Annual air intake = 8000 m3 per year (adult) Fractional portion of the year = 0.0356 (312 hours) I-131 ~ 4.57 E-14 pCi/m3 I-133 1.23 E-12 pCi/m Mn-54 ~ 2.39 E-15 pCi/m Sr-89 ~ 1.24 E-13 pCi/m H-3 ~ 4.62 E-10 pCi/m NMP2 Vent: Unit 2 average vent flowrate = 108.93 m /sec
- X/g value = 2.8E-06 (annual historical average)
Inhalation dose factor = Table E-7 of Regulatory Guide 1.109 - Annual air intake = 8000 m3 per year (adult) Fractional portion of the year = 0.0356 (312 hours) Co-58=4.87 E-14 Cr-51~1.81 E-12 H-3=4.97 E-10 The inhalation dose to a member of the public as a result of activities inside the site boundary is 1.18E-17 mRem to the thyroid (maximum organ dose) and 7.59E-18 mRem to the whole body. The dose from standing on the shoreline to fish is based on the methodology in the Off-Site Dose Calculation Manual as adapted from Regulatory Guide 1.109. During 1987, it was noted that fishing was performed from the shoreline on many occasions although waders were also utilized. In order to be conservative it is assumed that the maximum exposed individual fished from the shoreline at all times. The use of waders, of course, would result in a dose of zero from this pathway. The shoreline sediment doses are not taken into consideration by environmental TLD data. The Off-Site Dose Calculation Manual equation basically gives the total dose to the whole body and skin from the sum of all the plant related radionuclides detected in shoreline sediment samples. The plant related radionuclide concentration is adjusted for background sample results, as applicable. The equation, therefore, yields the whole body and skin dose by multiplying the radionuclide concentration adjusted for any background data (as applicable), times a usage factor, times the sediment or soil density in grams per square meter (to a depth of one centimeter) times the applicable shore width factor, times the regulatory guide dose factor, times the fractional portion of the year over which the dose is applicable. In order to be conservative and to simplify the equation, no radiological decay is assumed since the applicable radionucldies are usually long lived.
-23
ATTACHMENT 2 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1987) NINE MILE POINT NUCLEAR STATION PP2 DOSES TO MEMBERS OF THE PUBLIC DUE TO THEIR ACTIVITIES INSIDE THE SITE BOUNDARY (Continued) The calculation utilized the following information: Usage factor = 312 hours. Density in grams per meter = 40,000. Shore width factor = 0.3. Whole body and skin dose factor for each radionuclide = Regulatory Guide 1.109, Table E-6. Fraction portion of the year = 1 (used average radionuclide concentration over total time period). Average Cs-137 concentration = 0.40 pCi/g. Average Co-60 concentration = 0.06 pCi/g. The total whole body and skin dose from standing on the shoreline to fish is 1.01 ~
~ E-02 mRem whole body and 1.18 E-02 mRem skin dose for the period.
Doses to members of the public relative to activities inside the site boundary from aquatic pathways other than ground dose from shoreline sediment/soil are not applicable. In summary, the total dose to a member of the public as a result of activities inside the site boundary from the direct radiation, inhalation and shoreline dose pathways is 3.59 E-01 mRem to the whole body and 1.18 E-17 mRem to the maximum exposed internal organ (thyroid). The dose to the skin of an adult is 1.18 E-02 mRem. These doses are generally a result of the operation of NMP2. A portion of these doses for the direct radiation pathway are attributable to the NMPl facility.
-24
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ATTACHMENT 3 SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (1987) NINE MILE POINT NUCLEAR STATION $12 RADIATION DOSES TO LIKELY MOST EXPOSED MEMBER OF THE PUBLIC OUTSIDE THE SITE BOUNDARY JANUARY DECEMBER Radiation doses to the likely most exposed member of the public are evaluated relative to 40 CFR 190 requirements. The dose limits of 40 CFR 190 are 25 mRem (whole body or organ) per calendar year and 75 mRem (thyroid) per calendar year. The intent of 40 CFR 190 also requires that the effluents of NMP1 as well as other nearby uranium fuel cycle facilities be considered. In this case, the effluents of NMPl, NMP2 and the James A FitzPatrick (JAF) facilities must be considered. Doses to the likely most exposed member of the public as a result of effluents from the site can be evaluated by using calculated dose modeling based on the accepted methodologies of the facilities'ffsite Dose Calculation Manuals or may, in some cases, be calculated from the analysis results of actual environmental samples. Acceptable methods for calculating doses from environmental samples are also found in the facilities'ffsite Dose Calculation Manuals. These methods are based on Regulatory Guide 1.109 methodology. Dose calculations from actual environmental samples are, at times, difficult to perform for some pathways. Some pathway doses would be estimated using calculational dose modeling. These pathways include noble gas air dose, inhalat,ion dose, etc. Other pathway doses may be calculated directly from -environmental sample concentrations using Regulatory Guide 1.109 methodology. Since the effluents from the generating facilities are low, the resultant gaseous and liquid effluent doses are anticipated to be low. In view of this, doses can be based on calculated data. Doses are not based on actual environmental data for 1987 with the exception of doses from direct radiation, fish consumption and shoreline sediment. In addition, in order to be conservative and for the sake of simplicity, it is assumed in the dose calculations that, the likely most exposed member of the public is positioned in the maximum receptor location for each pathway at the same time. This approach is utilized because the doses are very low and the computations are greatly simplified. The following pathways are considered,
- 1. The inhalation dose is calculated at the critical residence because of the high occupancy factor. In order to be conservative, the maximum whole body and organ dose assumes no correction for residing inside a residence.
- 2. The milk ingestion dose is calculated utilizing the maximum milk cow location. As noted previously, in order to be conservative and for the sake of simplicity, the likely most exposed member of the public is assumed to be at all critical receptors at one time. In this case, the member of the public at the critical residence is assumed to consume milk from the critical milk location.
-25
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Attachment 3 (Cont'd) The maximum dose from the milk ingestion pathway as a result of consuming goat's milk is based on the same criteria established for item 2 above (ingestion of cow's milk). The maximum dose associated from consuming meat is based on the critical meat animal. The likely most exposed member at the critical residence is assumed to consume meat from the critical meat animal location. The maximum site dose associated with the consumption of vegetables is calculated from the critical vegetable garden location. As noted previously, the likely most exposed member of the public is assumed to be located at the critical residence and is assumed to consume vegetables from the critical garden location. The dose as a result of direct gamma radiation from the site encompasses doses from direct "shine" from the generating facilities, direct radiation form any over head gaseous plumes, plume submersion and from ground deposition. This total dose is measured by environmental TLD. The critical location is based on the closest year round residence from the generating facilities as well as the closest residence in the critical downwind sector in order to evaluate both direct radiation from the generating facilities and gaseous plumes as determined by the local meteorology. During 1987, the closest residence and the critical downwind residence are at the same location. The measured average dose for 1987 at the critical residence was 68.8 mRem. The average control dose (average of three locations) was 64.2 mRem. The average dose at the critical residence is greater than the average control location dose. A, major portion of this net dose is due to the differences between doses from naturally occurring radionuclides in the soil and rock at the different locations. This difference in dose rate can be demonstrated by observing the 1987 average dose for an environmental TLD located near the critical residence TLD but approximately 700 feet closer to the generating facilities. The annual average dose for this TLD location was 63.0 mRem. The dose for this location is lower than the critical residence location even though the TLD location with the lowest dose is closer to the generating facilities. The dose, as a result of fish consumption, is considered as part of the aquatic pathway. The dose for 1987 is calculated from actual results of the analysis of environmental fish samples. For the sake of being conservative, the average plant related radionuclide concentrations were utilized from fish samples taken near the site discharge points. The average concentration was adjusted to account for any background concentrations using, average control sample data. Only Cs-137 was detected during 1987 (net concentration was 1.9 E-03 pCi/g wet). The calculated maximum adult organ dose was 4.3 E-03 mRem to the liver. The maximum whole body dose is 2.8 E-03 mRem to an adult.
-26
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Attachment 3 (Cont'd)
- 8. The onl other a uatic P athwa Y considered is the shoreline sediment pathway relative to recreational activities. Doses from any other possible aquatic pathways are either zero or insignificant.
Environmental samples (2) collected during 1987 showed only naturally occurring radionuclides. Since no plant related radionuclides were detected, the calculated dose is zero. In summary, the maximum dose to the most likely exposed member of the public is 7.3 E-Ol mRem to the infant thyroid (maximum organ dose) and 2.4 E-02 mRem to the whole body (child). The maximum organ and whole body doses were a result of gaseous effluents. Doses as a result of liquid effluents were secondary, as noted above. The direct radiation dose to the critical residence from the generating facilities was insignificant or zero. These maximum total doses are a result of operations at the Nine Mile Point Unit 1, Nine Mile Point Unit 2 and the James A FitzPatrick facilities. The maximum organ dose and whole body dose are below the 40 CFR 190 criteria of 25 mRem per calendar year to the maximum exposed organ or the whole body, and below 75 mRem per calendar year to the thyroid.
-27
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TEMPORARY CHANGE NOTICE -? This temporary change shall be documented and approved by The General Superintendent-Nuclear Generation based upon recommendation of SORC members within 14 days in accordance with Technical Specification 6.8.3. TO: STATION SUPERINTENDENT, UNIT The attached Temporary Change was made to Procedure No. Rev. Title MODIFICATION RELATED C ~ CHANGE 0 YES g'NO Reason A o F6. MOD CONTROL NUMBER And ks.'recommended to be: ONE TIME ONLY.......... 0 ZN.EPPECT UNTIL NEXT REVISION The intent of the original procedure is t alter d Author Signature naiades'~ ~~ The temporary procedure was approved by: Supt.'evision Dept. Supv. Sig a'ture Date SRO Signature Date Station Date SORC MEMBER RECOMMENDATZONS (Minimum 2 regular members, 2 alternates)
- 1. 2 3 4 Recommend full SORC committee review this temporary change 0 0 0 0 Recommend Approval this temporary change does not change the intent of the original procedure and does not involve an unreviewed,safety question.
SORC r. Signatures Da e or SORC Meeting 1 I number (if required) 2 P 3 / 7
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~ g A GENERAL SUPERINTENDENT'(or designee) APPROVAL The temporary change is approved in accordance with Technical Specification 6.8.3. c Signatur Date FIGURE 2.0.5 SH 1 OF 1 hP-2.0 -32 May 1987
Offgas Noble Gas Detector Alarm Setpoint It Radiation Detector is sodium iodide crystal. itis The . a a scintillation device and has a thin mylar window so that is sensitive to both gamma and beta radiation. Detector response 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 se tpoint initially, assuming the nominal response lis ted on Tabl e 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 i Ci/ml cpm/>Ci/ml cpm B C D (ci) (cFi) ( Ci*CFi ) KR83 8.74E-2 KR85 4.90E-4 . 4.30E3 2. 11 KR85M 1.56E-l 4.80E3. 7.50E3 KR87 5. 23E-'1 8.00E3 4.18E3 KR88 5. 32E-1 7.60E3 4.04E3 F.R89 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 1.71E4 XE138 1.93 7. 10E3 1.37E4
.-AR41 ALS 4.99E4 The Offgas Noble Gas Monitor Alarm Setpoint equation is:
Alarm Setpoint ~ 0.8*350,000~2.1E-3*Xi(ci*CFi)/[f*zi(ci)]+ Bkg. Where the Alarm Setpoint is in cpm, Ei(ci*CFi) is 4.99E4 cpm, f is 25CFM and ZL(ci) is 8.36 pci/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 monito r. 7V y~~F~SIM ~ 44/A f-~ l~ + ~ C '~M Q$ "~ ~
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3 C TEMPORARY CHANGE NOTICE This temporary change shall 'e documented and approved by The General Superintendent-Nuclear Generation based upon recommendation of SORC members within 14 days in accordance with Technical Specification 6.8.3. TO: STATION SUPERINTENDENT, UNIT Z The attached Temporary Change was made to Procedure No. Q '9 CM Rev. Title f~ Crs CA4c sP Reason ID r) <tg~ I~ MrYf'<'fl Z~f k rO~d e n'+1A" m f'~ig=< f7ON Cvtfh. 7~+/'t)/~W 5 And is recommended to be: ONE IN TIME ONLY ~ ~ ~ ~ ~ ~ ~ ~ ~ EFFECT UNTIL NEXT REVISION,
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0 Th intent of the original proceduze is no altered. Author Signature D te wb'> The temporary procedure revision was approved by. Dept. Supv. Signat~ Da te SRO Signature Date S tation Supt. ~ Date SORC MEMBER RECOMMENDATIONS (Minimum 2 regular members, 2 alternates) 1 2 3 4 Recommend full SORC committee review this temporary change 0 0 0 0 Recommend Approval this temporary change does not change the intent of the original procedcre aad does not involve an unreviewed safety question. SORC Membe ignatures Date or SORC Meeting 1 2 number (if required ) 3 4 GENERAL SUPERINTENDENT (or designee) APPROVAL The temporary change is approved in accordance with Technica1 Specification 6.8.3. Signature Da e FIGURE 2.0.5 SH 1 OF 1 AP-2.0 -32 March 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's 10,200 gpm.. Bec'ause of tharge 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
'were conti'nue substantially even if different'rom the'istribution of the grab sample nuclides .in the tank obtained prior to discharge which 'was used to establish the detector alarm point. To avoid .this possibility of "Non . representative Sampling" resulting, in 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 355 cpm above background will be used until grab sample analysis with the required LLD sensitivity on TS Table 4.11-1 detects activity 355 cpm i s the same nominal setpoint as forThese the service are all water and cooling tower blowndown radiation'onitors. identical detectors.
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s THHPORhRT CELWE NOTZCE This temporary change shall be documented and approved by The General Supexintendent-Nuclear'eneration based upon recommendation of SORC members within 14 days in accordance with Technical Specification 6.8.3. TO STATION SUPERINTENDENT, UNIT pC The attached Temporary Change was made to u~ Procedure No. ODCW Title cg ~ 'ref 'rru
~ Rev. Z
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p+ <) Reason ;'n IC CCC Sr e uc((d a Om Z
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rN~n7'nd's recommended to be: ONE TIME ONLY ~ ~ ~ ' ~ ~ ~ ~ ~ ~ 0 IN EPPECT UNTIL NEXT REVISION ~ ~ The intent of the original procedure is no altered. Author Signature pate 3 ai 3'7 The temporary proceduxe revision was'approved by: Dept. Supv. Signature Date > 24/0 7 SRO Signature Date Station Supt. Date SORC MEMBER RECOMMENDATIONS (Minimum 2 regular membexs, 2 alternates) 1 2 3 4 Recommend full SORC committee review this temporary change 0 0 Recostteed dpproval thds tenporarf change does eet change the 1ntent of tha orlg1nal procedure and does not involve an unreviewed safety question. l~ 2 3g ~4 SORC Membe gnatures Date or SORC Meeting
~C- number (if required) 2 3
r GENERAL SUPERINTENDENT (or designee) APPROVAL The tempoxary change is appxoved in accordance with Technical Specification 6.8.3. Signa re Date FIGURE 2.0.5 SH 1 OP 1 AP-2.0 -32 Harch 1986
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TABLE 2-2 Ai~ VALUES - LIQUID* mrem ml hr- ui NUCLIDE 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-l 3.37E -1 Cr 51 1.28 3.21E2 2. 81E-1 7.63E-1 1.69 C'v Jog 'I 7Q 5'.$ 7c< /.0/t I '2.S'fC I Mn 54 8.36E2 1.34E4 4.38E3 1.30E3
&8 Fe 59 9.40E2 8.18E3 1.04E3 2.45E3 6.85E2 s(z(Z' Co 58 2.01E2 1.82E3 9.00E1 Co 60 5.70E2 4.85E3 2.58E2 >gb7 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. 9 1E-2 2.77E2 2. 72E-1 8.74E-2 1.37E-1 Hq ~/o I.'i& C( P. >~C3 /./o Ez l. /d/=z. Mo 99 2.05E1 2.50E2 1 o08E2 2.44E2 d~ ~" g,bjEz + old" go/8'2, y griE=2 +d )88 Y>/CZ YdK'2 I 131 1.26E2 5. 80E1 1.54E2 2.20E2 3.77E2 7.21E4 n/'5 7. S3 9,'igg 2. /.2788 /,mF/ I 133 2.78El 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 1.41E1 4.45E2 2.16E2 2.71E-l 9.22E-2 1.57E-1 Ce 141 2.48E-3 8.36El 3.23E-2 2 '9E-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-l 4. 18E-1 Calculated in accordance with NUREG 0133, Section 4.3.1
<9 May .1986
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TZllPQRARf CHAWE NOTICE This temporary change shall be documented and approved by The General Superintendent-Nuclear Generation based upon recommendation of SORC members within 14 days in accordance with Technical Specification 6.8.3. TO: STATION SUPERINTENDENT 3 UNIT The attached Temporaxy Change was made to Pxocedure No. &D u~ff Rev. Title (gart n~~/ C Reason CvS rS<rd':n rf d3A~ f ppan~~SI CdaP Zr Se >aO <Il Pftfff. And's recommended to be: ONE TIME ONLY . ~ . ' IN EPPECT UNTIL NEXT REVISION
~ ~ ~ ~ ~ 0 The intent of the original procedure is no altered.
Author Signature Date The temporary procedure revision waN'approved'by: Dept. Supv. Si ture D te WV~. SRO Signature Data Station Supt. Date 0 SORC MEMBER RECOMMENDATIONS (Minimum 2 r gular members, 2 alternates) 1 2 3 4 Recommend full SORC committee review this temporary change a a a a Retotttend dpproval - this teapotazT change does aot thaage tha latest of ths otdgdnal protadnra and does not involve an unreviewed safety question. lg ~2 3
~4 SORC Member gnatures Date or SORC Meeting QA Dmll number (if required) 2 3
GENIAL SUPERINTENDENT (or designee) APPROVAL The pox change is ap roved in accordance with Technical Specification 6.8.3. Si PIGURE 2.0e5 SH 1 OP 1 AP-2.0 -32 March 1986
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deciy constant for removal of activity on by, weathering, 5.73, x 10 .sec ".,'~fg leaf,;~i"'lant,'.:surfaces a e m '( corresponding to a,l4 day half-time) the transport'ime -.from pasture to cow,,to, milk', ;to";.
. 9gip receptor,.in sec.
'h the transport time from pasture, to harvest, to cow,'o,"'"~,.
milk, to receptor~ in sec. '~m,' f P ma fraction of the year that the cow is on pasture (dimensionless). fs em 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. cattle are considered to fed from two potential sources,
'ilk be pasture grass and stored feeds. Following the development 'kn Regulatory Guide 1.109, the value) of ~-ac@ fs will be I-"I-Pi
~considered unity SI) ~'rI!
Tabulated 9'~2 <9 ~aeon. below are dpd co ~5&v'eg +a ~ O.~ fo~ am, a~~er the appropriate parameter values aod their reference to Regulatory Guide 1.109. In case that the milk animal is t//ljP a goat, rather than a c'ow, refer to Regulator'y Guide 1-109 for the appropriate parameter values. .lg Parameter Value Table 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 iDVLL)a imrem/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) 7.78 x 10 (90 days) E-15 E-15 th (seconds) Qp (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]: Ri K'KaeFgpVapFLi[0 75(0-5/H)] (mrem/yr per UCi/m ) May 1986
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