ML17341B373
| ML17341B373 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 10/31/1981 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML17341B371 | List: |
| References | |
| PROC-811031, NUDOCS 8209160135 | |
| Download: ML17341B373 (74) | |
Text
{{#Wiki_filter:OFFSITE DOSE CALCULATION PANUAL FOR GASEOUS AND LICPJID EFF U""NTS FROH THE TURKEX POINT PLANT UNITS 3 Gc 4 F1or ida Power and Light Conpany October, 1981 8209160135 8209i0 PDR ADOCK 05000250 P PDR
'I lr / ~
OFFSITE DOSE CALCULATION MANUAL FOR GASEOUS AND LIQUID EFFLUENT 1.0 Introduction 2.0 Liquid Fffluent 2.1 Radioactivity Concentration In Liquid Paste 2.2 Radioactivity Concentration in Vater at the Restricted Area Boundary 2.2.1 Aqueous Concentration 2.2.2 Batch Release 2.2.3 Continuous Release 2.3 Method of Establishing Alarm and Trip Setpoints 2.3.1 Setpoint for a Batch Release 2.3.2 Setpoint for a Continuous Release 2.4 Accumulated Personal Dose 2.5 Projected Personal Dose 10 3.0 Gaseous Effluent ll 3.1 Introduction 0 3.2 Radioactivity in Gaseous Effluent 3.3 Effluent Noble Gas Monitor. Alarm Setpoint 12 3.4 Noble Gas Ga'mrna Radiation Dose Accumulated in Air 13 3.5 Noble Gas Beta Radiation Dose Accumulated in Air 3.6 Dose Due to. Iodine and Particulates in Gaseous Effluents 17 3.7 Dose to a Person from Noble Gases 22 3.7.1 Gamma Dose to Total Body 22 3.7.2 Dose to Skin 3.8 Projected Doses due to Gaseous Effluent 23 4.0 Dose Commitment from Releases over Extended Time , 25 4.1 Releases during,a Quarter 25 4.2 Releases during 12 Months Appendix A Pathway-Dose Transfer Factors Appendix B Nuclide Distribut'on in the Cooling Canals B-l Append'x C Limited Analysis Dose Assessment for Liquid C-1 Radioactive Effluents Appendix D Technical Bases for A D-1 Appendix E Technical Bases for Eliminating Curie Inventory E-1 Limit for Gaseous Paste Storage Tanks Appendix F Base .of Techn'cal Specification 3. 9. l. d. 1 Limit's
is 0 0
OFFSITE DOSE CALCULATION MAi~lVAL FOR GASEOUS AN D LIQUID EFF LUENT
- l. 0 Introduction This Manual describes acceptable methods of calculating radioactivity c'oncentrations in the environment and the potentially resultant, doses+ 1 of+site~ that are associated with liquid and gaseous effluents from the Turkey Point.'Nuclear Plant. The radioactivity concentrations and dose estimates are used to demonstrate compliance'with Technical Specifications required by 10 CFR 50.36. The methodology stated in this Manual is acceptable for use in demonstrating operational compliance with 10 CFR 20.106, CFR 50 Appendix I,'nd 40 CFR 190. Only the dose attributable to the Turkey Point Units 3 and 4, is considered in demonstrating compliance with 40 CFR 190 since no other nuclear facility exists within 50 miles of the Plant.
Monthly caizculations are made to guide the management of station effluents and.to veri y that potential radioactivity "concentrations and doses offsite satisfy the Technical Specifications. The receptor is described such that the exposure of any resident near the plant is unlikely to be underestimated. Even more conservative conditions (e.g. location and/or exposure pathways expected to yield higher computed doses) than appropriate for the maximally exposed person may be assumed . when calculating the concentration or dose. Monthly calculations made to assure that a'ir dose and dose commitment specifications are'ot exceeded are based on atmospheric dispersion and k deposition of gaseous effluents derived from reference meteorological conditions.*+>> Calculations made to assess the radioactive noble gas dose to air are based on the location offsite that could be occupied by a person where the maximum air dose is expected. Dose is commonly used to mean personal dose equivalent commitment. Offsite means outside the exclusion area. Reference meteorological conditions are annual aver ged conditions during years 1976 and 1977.
t ~ 0 e
~ I page 2 Calculations of dose committed from rad'oactive releases over extended time (3 and 12 months) are also made for the purpose of verifying compliance with regulatory limits on offsite dose. For these. calculations the receptor is selected on the basis of the combination of app pathways identified in the land use census and the maximum
'cable'xposure ground leve1 X/Q at a residence, or on the basis of more conservative
'conditions such that the dose to any resident near the. Plant is~unlikely be underestimated.
'to
~ 2.0 Liquid Effluent
- 2. 1 . Radioactivit Concentration ln Liquid Waste The concentration of radionuclides in liquid was=e is determined by sampling and analysis in accordance with Table 3.9-1 of the Technical Specifications. When a radionuclide concentration is below the lower-limit of detection (LLDE foz the analysis, it is not reported as being present in the sample.
I
- 2. 2 Radioactivit Concentzation in Water at the Restricted Area Eoundar Technical Specifications 3.9. l.a.l and 3.9.1.a.2 require measured radioactivity concentrations in liquid releases to. be used to calculate.
the fraction of the unrestricted area maximum pe'zmissible concentration (MPC) (10 CFR 20; Appendix 8, Table 2r Column 2). For compliance with these specifications, tne restricted azea boundary is. assumed to be at tne end of the condenser cooling water mixing basin where water enters the system of cooling canals. Radioactive material in liquid effluent is diluted by condenser cooling water from fossil units 1 and 2 and from:nuclear units 3 and 0. When estimating the unrestricted area activity concentration in water, the total condenser cooling water flow into the condenser cooling water P
~ IJ, mixing basin from operating condenser cooling water pumps at the. four units is assumed for di1ution.
2.2.1 Aqueous Concentration. The diluted concentration of radi'o-nuclide i in the condenser cooling water mixing basin and its outflow is estimated with the equation C .'.= zi C.. i F
-1 F
~
i page 3 where C. i concentration of radionuclide i in liquid radwaste released 0 C zi (pCi/ml) concentration of radionuclide i in the water in the condenser cooling water mixing basin and its outflow (pci/ml) Fl/P2 dilution F flow in radioactive 1'quid discharge. line (galdmin)~- P F = minimum total condenser cooling water flow (gal/min)
- Value not greater than the rated total condenser cooling
'ater flow from operating condenser cooling water pumps at the four units.
2.2.2 Batch Release. A sample of each batch of liquid radwaste is analyzed before release fo'r E-131 and other principal gamma emitters, or for total activity concentration. The fraction of the unrestricted area MPC present in the batch tank, r~~.B b is derived either from isotopic analyses or from gross B-'Y analysis. With the activity concentration in a'batch sample based on cumulative or gross P-'Y activ'ty alone, the fraction o the unrestricted area MPC in the batch is estimated by Fi&" = b ~b C (2) 3 x10-where Cb bp
= activity concentration in batch sample measured by gross P-Y analysis or cumulative total of isotopic analysis (pCi/ml)
-8 3 x 10 = unrestricted area HPC for.unidentified radionuclides in water (pCi/ml) when the fraction of the unrestr'cted area MPC is
'lternately, derived from timeisotopic analyses identifying ioding end principal gemma'-'
emitters> FilBb is calculated with the equation I PiB C (.3) b bpi 0 I HPC. i identified
*F and 1
F 2 may have any suitable but identical units of flow vo 1 ume/ ).
II
~
t% I 4 page 4
\
where C bpi.
= concentration of radionuclide' (including T-131 and principal P
gamma emitters) in a batch sample measured prior to release (pCi/ml.) HPC. = activity concentration limit in water of radionuclide i according to 10 CFR 20, Appendix B, Table 2, column 2 (pCi/ml) Quarterly average of the fraction of HPC in the batch tank due to E-131 end orincioel amma emitters Quarterly average of the fraction of HiPC in the batch tank due to ell radionuclides measured E E b is an adjustment to account for redionuclides not measured prior'.to release but measured in the monthly end ouerterly sample'per Technical Specification Table 3.9-1. "The value of Zb has been determined based on past operating data end is E = 0 8
,b The fraction of the unrestricted area MPC present in the condenser cooling water mixing basin outzlow due to a batch release, FHPCb mey be calculated with the equation PiiPC MB . F (4)
F where FHB b fraction of the unr'estricted area HPC present in the batch tank from equation (2) if cumulative or, gross ectivity if used or from equation (3) iz isotopic analysis is used. flow in the batch release line ( gel/min).~ 'alue not less than the rated or measured pumping rate through the batch discharge line.
*F end F may have any suitable but identical units of low 1 2 (volume/time).
~ c%
page 5 2.2.3 Continuous'elease. Continuous aqueous discharges are sampled and analyzed according to the schedule in Technical Specifications Table 3.9-1. The fraction of the unrestricted area MPC present in a continuously discharged radioactive stream, FC ., c is derived either from isotopic analyses or from gross P-Y analysis. With the activity concentration in a continuous radioactive release stream based on the cumulative or gxoss P-Y activity alone, the fraction of the unrestricted area MPC in the waste stream i.s estimated FC = C c I c 3 x 10 where C c
= activity .concentration in continuous release measured by gross P-Y analysis or cumulative total of isotopic analysis (pCi/ml )
Alternately> when the fxaction of the unrestxicted area HPC is derived from isotopic analyses, FC., is estimated with the equation ~ . FC c
- Ec
- i. i identified concentration of radionuclide i (including l-131
~
'rincipal whexe C cwi,
=
gamma cmitters) measured in weekly continuous discharge stream sample'f and Quarterly average fraction o+.HPC due, to I.-131 and principal gamma emitters measured in weekly samples of continuous releases durin the uarter Quarterly average fraction of HPC due to a'1 radio-'. nuclides measured in samples of continuous releases E c is an adjustment to account or xadionuclides measured in monthly and quarterly composite samples but not in weekly samples of continuous releases. The value of ic nas been detexm ned based on past operating data and i.s E = 0 9 c
0 4 > page 6 The fraction of the unrestricted area=HPC present in the condense= cooling water basin outflow due to continuous aqueous discharges, PMPC may be est imat ed with .the equation c > FHPC = FC . F (7) F where FC c
= fraction of unrestricted area ~>iPC present in each
'continuously discharged radioactive stream from equation (S) if cumulative or gross activity use or rom eouation (6) if isotopic analysis used.
F lc
= flow of continuously released aqueous radioactive discharge ( gal/mi:n) 2.3 Hethod of Establishin~ Alarm and Trio set pints The alarm/trip setpoint for each liquid effluent radiation monitor is derived from the concentration limit provided in 10 CFR Part 20, Appendix B, Table 2, Column 2 applied in the condenser cooling water mixing basin outflow. That is, 'the alarm .setpoint is based on a concentration limit in the mixing basin outflow. Radiation monitoring and isolation poi'nts are located in the steam generator blowdown 1'ne, R 19> and the liquid waste disposal system line) R 18>', through which radioactive waste effluent \ is eventually discharged into the canal basin.
The alarm setpoint for each liquid e fluent monitor is based upon the measurements performed according to Technical Specificati.on Table 3 9-1,
. of radioactivity in a batch of liquid to be relea'sed or in the continuous aqueous discharge. Alternately> the alarm setpoint may be based upon or gross 8-Y activity of the liquid waste. ~'
'umulative
\
page '7 ~ ~ 2.3.1 Set oint for a Batch Release. The liquidradwaste effluent line radiation monitor'alarm setpoint is determined with .the equation S A . F2 PiJB ls or a method which gives a lower setpoint value.. where S = radiation monitor alarm setpoint (cpm) counting rate (cpm/ml) or activity concentrat'ion (+Ci/ml) in laboratory of sample from batch tank
= ratio of effluent radiation monitor counting rate to laboratory counting rate or activity concentration in a given batch liquid (cpm per cpm/ml or cpm per pCi/ml)
~ F = flow .in the batch release line (gal/min).~ Value not less ls than the rated or measured pumping rate through the batch discharge line.
F = minimum total condenser cooling water flo'w (gaI/min).* Value not greater than the rated total condenser cooling water pumping rate from pumps operating at the 0 units. FHB b
= is determined as described in g 2.2.2.
2.3.2 Set@oint for a Continuous Release. The alarm setpoint of the radiation monitor on a continuous radioactive discharge line is determ ned with the equation S A . F . g FC F c lc 1 or by a method which gives a lover setpoint value.
~
where A = activity concentration (pCi/ml) or counting rate (cpm/ml) in laboratory of weekly sample. F flow in the radioactive liquid continuous discharge line (ml/sec).* Value not less than the rated or measured pumping rate through the discharge line I F total condenser cooling water flow (ml/sec).~ Value not greater than the rated total condenser cooling, water pumping rate from operating pumps at the 0 units. FC = is determined as described in g 2. 2. 3. c
*F and F may have any convenient units of flow .(i.e, volume/time) 1 2 provided r the units are dentical'.
k ~ <
~ q page 8 2.4 Accumulated Personal Dose l
Technical Specification 3.9. l.bP requires the dose or dose commitment it to a person oz s te d uee to radioactive material released in liquid effluent to be calculated on a cumulative basis at least once every 31 days. Tne requiremen tss is satiszied by evaluating the accumulated dose commitment to a hyp'othetical adult exposed by eating fish and shellfish taken from the cooling canals. The model that is used to evaluate doses due to radioactivity in liquid ezzluents is shellf ish (10) i i nk V where D nk
= the dose comm'tment (mrem) to organ n due to the radio-nuclides identified in sample analysis k where the analyses are those required by technical specizications Table 3.9-1.
(The contribution to the dose from gamma emitters become available on a batch basis for batch releases and on a' weekly basis for continuous releases. Similarly the contributions from H-3 are available on a monthly'asis and the contributions from Sr-89 and Sr-90 become available on a quartex ly basis.) A. in transfer factor relating a unit aqueous concent'ration o raoionuc 1 iae -i (pCi) to dose commitment rate to organ n or total body of an exposed person tabulated in the ODCH Appendix A (mrem/hr per pCi/ml) C. ik
= the concentration of radionuclide i in the undiluted liquid waste to be discharged (pCi/ml')
tk = period of time (hours) during'nich liquid waste represented by analysis k is discharged . F = liquid waste discharge flow during release represented by lk simple k (gal/min)
'i = effective decay constant (minute ) for nuclide i>(~ .
where X. is 'the radioactive decay constant F
'3 canal-ground watex interchange. ~low> approximate y 2.2 x 10
; gal/min
= cooling canal effective volume> approximately 3.75 x. 10 gallons
~ ~ I ~ ~ k Summation over releases represented by the various samples ana1.yzed gives the total dose to each organ D = g' k Vhere n the dose commitment to organ n, including total body, of the maximally exposed person during the quartex" to da.te (nrem) For the quarterly dose assessments to be included in the Annual Radiological Environmental Monitoring Repoxt required by Specification 6.9.4.b, doses will be calculated with eouations (10) and (ll) for all age groups and organs on the basis of radionuclides measured in liquid radioactive effluent according to the sampling and analyses required
\
in Technical Specification Table 3.9-1. Based on an evaluation of the radionuclde distribution typical in liquid radioactive effluents, the calculated doses to individuals have been determined to be dominated by the radionuclides, Co-58, Co-60, Nb-95, Ag-110m', Cs-134, and Cs-137. These 6 nuclides typically contribute 907. ox more of the adult's total body dose and the adult's GI-LLT. dose, which is the critical organ and critical age group'.'herefore, the dose commitment due to radioactivity in liquid effluents may be reasonably evaluated by limiting the dose calculational process to these radionuclides for the total body end the GX-LLI.,'quationi (10) can be simplified in. the following ways. pathway ) and the ef~ective decay constan t
- 1) The transfer factor (A . C pathway
(),.) can be combined into a single. factor A . 1 1
- 2) The concentration (C..k), liquid discharge flow (Flk), and time of the release (t ) can be combined to provide the total release for a specified period of time (eg> Ci/calendar quartex).
page 10 These combinations pxovi.de the following simplified equation D = 0.8 1
. V i,'n Afish Ashell ish in ( 12) where Q i = ofthe total release of radionuclide i for the soecified period time, e> .monthly 0.8 = a conservatism factor to a.liow for variability in radionuclide distribution (only include if using simplsfied approach).
Refer to Appendix C for a detailed evaluation end explanation of this simplified approach. Monthly dose calculations required by Specifications 3.9.2.b.l and c., I 3.9.2.~ mey be done with either equations 10 and 11 or equation'2. Vhen equation 12 is used, the doses should be calculated for the adult total body end. adult GX-LLE from the cumulative release for:the radionuclides Co-58, Co-60, Hb-95, Ag-110m, Cs-134> and Cs-'137. 2.5 Pro ected Personal Dose Technical Specification 3.9.l.d.l requires the doses to a person offsite due'o radioactive material released in liquid effluent to be projected over a quar'ter et least one time during each month. This requirement is satisfied by calculating the total body and organ dose commitments to a hypothetical adult exposed by eating fish and shellfish taken from the cooling canals. Section 2.4 includes the method for doing this calculation.. Appendix P pxesents the technical bases for using this projection to determine liquid radwaste equipment operation. The dose from liquid offluents is ~ro'ected by eztrspols ing the dose commitr,".ent to date during the current quaxter to include the entire quarter. On the basis of total activity released, dose commitment is projected with the relation P=91.D (13) X where P = the projected total body or critical organ dose commitment .to the maximally exposed person (mrem) 91 = number of days in a calendar quarter '
= number of days in current quarter represented by available radioactive dischaxge data
page>> 3.0 Gaseous Effluent
- 3. l Introduction Units 3 and 4 discharge gaseous effluent through the plant vent, Unit 3 Spent Fuel Pit vent> air ejector vents, and steam gene, ztor blowdown vents. These gaseous effluent streams, radioactivity monitoring points,,and effluent discharge, points are illustrated schematically in Figure 3-1.
3.2 Radioactivit in Gaseous Effluent ' For the purpose of estimating offsite radionuclide concentrations ' and radiation doses, measured radionuclide concentrations in gaseous t effluent and in ventilation air exhausted rom the Plant are -'elied upon'.- Table 3.9-3 in the Technical Specifications identif es specific radio-nuclides in gaseous discharges for which sampling and analysis is done. When a radionuclide concentration is below the LLD for the analysis, it is not reported as being present in the sample.'oble Gases. T'ne distribution of radioactive noble gases in a, gaseous effluent stream is determined by gamma spectrum analysis of identifiable radionuclides in effluent gas sample(s) Results of one. 'r more previous analyses may be averaged to obtain a representative spectrum. In the event the distribution is unobtainable from measured data; the distribution of radioactive noble gases appearing in. Table 3-, 2 herein may be assumed. Some gaseous effluents from both Units. 3 and 4'. whose sources are identified in Table \3-1> discharge in common through the Plant Vent. Mhen needed to assure that the effluents are within allowable limits on a per reactor bases; the measured release from the Plant Vent is apportioned to each unit on a ratio equal to the ratio of primary coolant radioactivity concentration in the two reactors during the quarter.. Vhen calculating atmospheric disper sion of gaseous ef fluent, gaseous discharges from Units. 3 and 4 are treated as a mixed mode release from a single composite vent.
pe page 12 3.3 Effluent Noble Gas Honitor Alarm Setooint Instrumentation is provided to monitor gamma radiation from radio-E active materials released from the Plant in gaseous effluents. ach monitor includes an alarm that is set to report at'r below the level at which radioactive noble gas .in gaseous effluent from a monitored
~ stack or vent exceeds a rate calculated to cause a noble gas concentration
'I offsite equal to that specified in 10 CFR 20> Appendix B, Table 2> Column 1
~
for the mixture. Setting effluent noble gas monitors to trigge an alarm at or below the concentration limit assures that action can be taken to ensure that the unrestricted area concentration specified by 10 CFR Part. 20. 1'06 and the corresponding dose rate limits in Specification 3.9.2.a are not exceeded. The gross activity concentration of noble gas correspond'ng to the 10 CFR Part 20> Appendix B> Table 2, Column 1 limit is'alculated from the distribution determined in 5 3.2 with the equation HPC=C g Ci i HPC i where MPC = gross activity concentration of noble gas mixture corresponding, to 10 CFR Part 20 Appendix B, Table 2 column 1 limit (pCi/cm 3 ) C. L
= activity concentration of noble gas radionuclide in gaseous e release (PCi/cm 3 )
C' P C. = activity concentration of noble gae mixtn-e releaaeg 3 (yCi/cm ) M3?G..' 1. 10 CFR Par't 20 Appendix B, Table 2, Column 1 value. Note that this is simply the aggregate of the concentrations of radio
'nuclides in a sample divided by the fraction or mult'pie of HPC constituted by radionuclides in the same sample. Por purposes of simplifying evaluations of HPC, the total activity concentration of th'
~ noble gases may be assumed to be Kr-88, which has the most restrictive HPC value (2 x 10
-8 pCi/ml) of the no'ole gases.
The alarm setpoint for the effluent noble gas monitor is then calculated with the equation S= HPC. h (lS) 4.7x10 . F. X Q
~ I ~ page 13 or by a method that gives a lower setpoint where S = alarm counting rate setPoint (cPm) or {mR/hr) h = corn pCi/cm 3
'r effluent noble gas monitor counting rate response calibration mR/hr for,noble gas pCi/cm 3 gamma xadiation discharge rate of gaseous effluent (ft3 /min)
X/Q atmospheric dispersion for mixed mode relea'se from ~ release point to unrestric" ed area (pCi/m3 per pCi/sec). 3
'4.7 x 10, = conversion constant 1 m . 1 min 35.31 ft3 60 sec
= maximum permissible concentration according to 10 CFR Part 20 > Appendix 3 > Table 2, Co lumn 1 (pC iIcm 3
)
The value of X/Q adopted in a setpoint calculation will be based eithex on prevailing meteoxological conditions or on re erence meteo-. ological conditions. Minimum atmospheric dispersion offsite derived from reference meteoxological conditions>> 'at the site boundary 1950 meters SSE of the Plant are: X = 5.8 x
'7,10 sec'/m 3 3.4 Noble Gas Gamma Radiation Dose Accumulated in Air Technical Specification 3:9.2.b requires that the offsite air dose from noble gas gamma radiation not exceed 5 mrad per reacto- during any 1
.calendar quarter. Specification 3.9.2.b.g requires an evaluation be performed monthly to verify that the accumulated air dose does not exceed the limit.
The quantity of radioactive noble gas. discharged duxing an interval of time is determined by integrating the release ra'te measurement oX 'each effluent noble gas monitor identified in Figure 3-1. The .total measured radioactivity discharged via a stack ox vent duringa counting interval is determined by the relation 3 N.. F (16)
'.53 x 10:. h
>>Reference meteorological data are tabulated in Tables 3.5, 3.6 Y
and 3.7. Their derivation is descr.'bed in Revised Radiolo~ical Eff1 uent Technical Specifications: Gaseous Effluent Dilution Factors, Florida
' 'S
page 14 where Q. = total measured gaseous radioactivity release vi.a a stack or j vent during counting interval j (pCi)
-5 3 "3 3.53x10. = conversion constant (ft /cm. )
N. = counts accumulated during counting interval j F = discharge rate of gaseous effluent stream (ft3 /min) h = effluent noble gas monitor calibration or counting rate s rasponsa for nobis gas gamma radiation corn
'3 s
pCi/cm The distribution of radioactive noble gases in gaseous releases is determined by gamma spectrum analysis of gaseous e=fluent samples in accord with Technical Specification Table 3.9~3. 'In the event the s radioactive noble gas distribution is not obtainable ==om sample(s) taken during the current period the distribution will be ootained from the most recently available data or from Table 3-2; If f. represents the fraction of radionuclide i in a -given'ffluent stream, then the quantity of radionuclide released in a given gaseous effluent stream during counting interval j is estimated by tne relation 4 17:) The gamma radiation dose to air offsite as a consequence o noble . gas discharged from each unit. can be calculated with the equation D= 1 0 8
.X.A.Q.
Q eff (18) where D = noble gas gamma dose to air due to effluent from mixed-mode release (mrad) A = r- a dionuclide distribution weighted (effective) factor converting time-integrated, ground-level, total act'vity concentration of radioactive noble gas to air dose due to s radiation
~
gamma mrad 3 {pCi . sec)/m s X/Q = atmospheric dispersion factor for a mixed-mode'ischarge (sec/m 3 ) 0.8 = a conservatism factor which, in effect, increases the estimated dose to compensate for variability in 'radionuclide
.distribution.
\
page 15 An effective gamma air dose factor> A> 3 has been derived from noble eff gas radionuclide distributions in routine operational releases. Refer to Appendix D for 'a detailed explanation. The effective gamma air dose factor derived is A '1.4x eff 10 mrad (>Ci . sec)/m 3 Alternately, the gamma air dose may be calculated with the equation D' X P Q Q.. f:A .(19) where D' = noble gas gamma dose 'to air due to effluent from mixed mode release (mrad) A . = factor converting time integrated, ground-level concentration of noble gas i to. air dose from gamma radiation, listed in Table 3-3 mead (pCi sec)/m 3 0 (X/Q) = atmospheric dispersion factor (sec/m 3 ) for a mixed mode Specification 3.9.2.b.g is satisfied by calculating the noble gas discharge gamma radiation dose to air at the o~fsite location'identified in Figure 3-2., At that location> 1950 meters SSE of the Plant; the refeience atmospheric dispersion factor to be used is X -7
= 5. 8 x 10 sec/m 3 3.5 Noble Gas Beta Radiation Dose Accumulated'n Air Technical Specification 3.9.2.b requires that the o fsite, air dose from noble gas beta. radiation not exceed 10 mrad during any calendar quarter. The quantity and radionuclide distribution of radioactive noble gases discharged as airborne effluents are determined as described in ~
J 3.4 herein.
~ ~
page 16 l Technical Specificat'ion 3.9.2.b. requires a monthly determination'f i whether cummulative noble gas releases cause a beta radiation dose i offsite to 'air i iin excess of the limit stated in Specification 3.9.2. o. This determination can be made by using the equation D = l . X. Ap. (20) 0 8 Q eff j where D
~
= nooblee ga gas beta dose to air due to a mixed-mode release (mrad)
A = radionuclide distribution weighted (effective) factor converting time-integrated, ground-)evel, total activity concentration of radioactive noble gas to air dose due to beta radiation -a (yCi . sec)/m 3J/ X/Q = atmospheric dispersion factor for a mixed mode discha ge
'3 (sec/m )
An effective beta air dose factor, A~, has been derived from noble 0 gas I eff rzdionuclide distributions in routine operational releases. Re=er to Appendix D for a detailed explanation. The effective beta air dose factor derived is As 3.4 x 10 ml ad (pCi . sec)/m 3 l Alternately> Specification 3.9.2.b.g may be satisfied by calcu ating 1 the beta radiation dose to air offsite with the equation D) = X Q.. f... Ap (21) Q j i where A i=
. factor converting time-integrated, ground-level concentration of noble gas radionuclide i to air dose from beta radiat on>
listed in Table 3-3. mrad (pCi sec)/m 3( g
I page 17 Specification 3.9.2.b. is satisfied by calculating,the noble gas beta radiation dose to air't the location identified in Figure 3-2. At that location, 1950 meter's SSE of the Plant, the reference atmospheric dispersion factor to be used is
-7 X =
5.8 x 10 sec/m 3 Q l&lVH) 3.6 Dose Due to Todine and Particulates in Gaseous Ef luents C ' ' Technical Specification 3.9.2P'equires tnat $ I-131 +Y'>+ lvv> I and radioactive material in particulate form-having a half-life 'greater than 8.0 days, in gaseous effluents released to the area offsite cause a dose to any organ or the total body of a member of
'the public no more than 7.5 mrem during a calendar quarter.
Radionuclides other than noble gases in gaseous effluents that are, measured by the radioactive gaseous waste sampling and*analysis program described in Technical Specification Table 3.9-3 are used as the release term in dose calculations. Airborne releases are discharged either via a stack above the top of the containment building or via building vents and are treated 'as a mixed, mode re1ease from a single location. For each . of these release combinations, samples are analyzed weekly> monthly, auarterly, or or each batch release- according to 'able 3 9-3. Each sample provides .a measure of the concentration of specific radionuclides, C i.> in 'gaseous effluent discharged at flow, F a , during e time increment 4t. Thus> each re1ease is quantified according to the relation ik ik " aj j '(22) where Q.k ik the quantity of radionuclide i released in a given. ef fluent stream based on analysis k (pCi) ik concentration of radionuclide i in gaseous ef f luent .identif ied 3 by analysis k (pCi/m ) F effluent stream discharge rate during time increment4 t. (m3 /sec) aj j 3 time increment j during which radionuclide I i at concentration ik is being discharged (sec) C.k
'page 18 A person may be exposed directly to an airborne concentration of i
radioactive material discharged in effluent and indirec" ly via pathways i involving deposition off radioactive material onto the ground. Dose estimates account for the exposure via applicable ones of the, following pathways-') direct radiation from noble gases
- 2) inhalation
- 3) direct radiation from ground plane deposition
- 4) 'fruits and vegetables
- 5) air-grass-cow-meat 6.) air-'grass-cow-milk Of cll these pathways>. the air-grass-cow-milk pathway is by far the controlling dose contributor. Of the dose by this pathway> the radioiodines contribute essentially all o the dose, with I-131 typically contributing greater than 95/. The dose transfer factors .for the radioiodines are much greater than any of the other radionuclides. The critical organ is the infant's thyroid. For this reason, the potent'al critical organ dose via airborne effluents can be estimated by simply determining an effective dose transfer factor. for the radioiodines based on the typical radioactive effluent distribution> the grass-cow-milk-man pathway,'nd
": the infant thyroid zs the receptor. Then for conservatism the total cumulative release of all radioiodines and particulates can be applied to the effective factor and a conservative estimate of the infant
'I t
'hyroid dose determined.
The requirement> in Specification 3.9.2.~ to determine montnly whether cumulative releases have caused a total body or organ dose C. commitment in excess of limits in Specification 3.9.2. may be met by using the .following. equation'. DMk= 3.17 x 10 . D . TG (2Z) 0.8 Q
page 19 where DM k
= the dose commitment (mrem) to an infant's thyroid received from exposure via the air-grass-cow-milk pathway attributable to iodines identified in analysis k of effluent. air D/Q = relative deposition rate onto ground from' mixed
-2 mode atmospheric release (m )
TG 131
= factor converting ground deposition of radxoiodin'es to the dose commitment to. an in ant's thyroid exposed via the grass-cow-milk pathway, mrem/ r yCi/(m2 . sec) 3.17xl 0 conversion constan~ (yr/sec) 0.8 = a conservative factor which> in effect; increases the estimated dose to compensate for variability in the radionuclide distribution.
When equation 24 is'sed to'stimate the criti'cal organ ( infant' thyroid) dose commitment, the effective dose transfer factor used in the equation is ( TG = 6.5 x 10 ll mrem/v-131 2 NCi/(m . sec) The reference data from which TG was derived are summarized in Tail'e D-2 131 of Appendix D. Alternately, the monthly determ'nation, required by Specification C. I 3.9.2.~ may be made by using equations 23, 24, 25 and 26,. Quarterly calculations of dose commitments due to radioiodine and radioactive particulates in effluent air for inclusion in the Annual Radiological Environmental Honitoring Report are performed using equations 23 through 26, following. The dose commitment via exposure to airborne concentrations resulting from a release, Q. , of airborne radioactive material other than noble is, calculated with the equation ik'as, ankk 3.17 x 10 X d
. Q
~ Q'kik 7
~ TA snip (24)
page 20 The dose commitment via exposure pathways involving radionuclide deposition from the atmosphere onto vegetation or the ground is calculated
-with the equation ankk=3.17x10 (25)
D D Q ik P where D ank the dose commitment (mrem) to organ n of a pe" son in age i s group a due to radionucliae identified in analysis k. of an air effluent where the analysis is required by Technical Specification Table 3.9-3. I TA . = a factor converting airborne concentration of radionuclide anip . i to dose commitment t'o. organ n of a person in age group a where exposure is directly to airborne .-..a"erial via pathway p (inhalation, or external exposure to the plume) ~nrem/ r
)ICi/m TG anip. = factor converting ground deposition of radionuclide i to dose commitment to organ n of a person in age group a where exposure is directly or indirectly to radioactive material s
'2 WCi/(m . SeC) o X /Q = atmospheric dispersion factor for a mixed 'ode release, adjusted for depletion by deposition (sec/m 3 )
D/Q = re1ative deposition rate onto ground from a mixed mode 1
-2 atmospheric release (m )
-8 =
3.17 x 10 conversion constant (yr/sec) I The concentration of tritium in vegetation is a function of the airborne concentration rather than the deposition. Thus the dose commitment from airborne H via vegetation (fruit and vegetables), air-grass-3 cow-milk, or'air-grass-cow-meat pathways is calculated with the equa~'ion D ankk 317 x10 . X. Q.ik I( 26/ p where X/Q atmospheric dispersion factor, for a mixed mode relea'se (sec/m 3 ) 3.17 x 10 conversion constant (yr/sec)
page 21 e The dose commitment discharges from a V via a given pathway as a result. of measured release point is accumulated with D an
=
~k.
D ank (27) The counting'ndex k may represent either p, analysis of a grab sample w, a weekly sample analysis m, a monthly composite analysis, or q, a quarterly composite analysis The maximum value of each counting index is the number of analyses in a sample category'. 'he total dose commitment during a period of time is obtained by summing the contributions via separate release points and via separate environmental pathways. ~ ~ h th do to p o d to 'od d p e' as airborne effluents is calculated as required by Specification 3.9.2.~ the air-grass-cow-milk is evaluated by assuming a cow on pasture 4.5 miles west of .the plant. (There is no milch or meat animal within 5 miles.) At that loc'ation, reference atmospheric dispersion and deposition factors are. X = -7 d 1 x 10 Q 3 The inhalation, fruit and vegetable, and iiradiation by airborne radxonuclides and'by deposition on the ground pathways are evaluated at the nearest garden
'd (with residence assume d) 3.6 . miles m west north~est of the plant. A.t th t location, reference atmospheric dispersion and depose~i'on factors a e.
' -7 i' x 10 sec 3
,Xd 9 x10 -8 Q m
-10 -2 D 5x10 m
page 22 1 3.7 Dose to a Person from Noble Gases Technical Specification'.9.2. requires the calculation of the dose or dose commitment to a person offsite exposed to 12 consecu" ive months of radioactive liquid and gaseous effluents from the plant. One component of personal dose is total body irradiation by gamma rays from nobl.e gases. Another is irradiation of skin by beta and gamma radiat'on =rom noble gases. The methods of calculating these doses are presented in sections 3.7.1 and 3.7.2. The amount of radioactive noble gas discharges is determined in the. manner described. in section 3.4. 3.7. l Gamma t Dose to Total Bod The gcn'nc, -zad" rat"o-.. des+ to the who1e body of a member of the. public as a consequence of .".ob.= "as released from the Station is calculated with the equation. D = Q.. X . PY. (28) where D Y
= noble gas gamma dose to total body (mrem)
P'. Yi
= factor converting time integrated, ground. level concentration of noble gas nuclide i to air dose from gamma radiation listed in Table 3-0 mrem (pCi sec)/m 3 When the total body dose due to'amma radiation from noble gas required by Technical Spec'fication 3.9.2.g'.2 r or 6.9.4.6 is calculated, the most exposed receptor is located 3.6 miles west northwest of the plant where the reference meteorological dispersion factor> X/Q, is
-7 3 1 x 10 sec/m .
t ~ page 23 3.7. 2 Dose to Skin ~ The, beta radiation dose to the skin of a member of the public due to beta radiation from noble gas released from the Plant may be calculated with the equation Dp ='j'. Q..' Q s where D~
= noble gas beta dose to skin (nrem)
~ s
'Sp'. =
Pi factor converting time integrated ground level concentration, of noble gas to skin dose from beta radiation listed in Table 3-b 'rem lLCi sec 3 m When the skin beta dose due to noble gas required by Specification 6.9.4.b is calculated, the most exposed receptor is located 3.6 miles west northwest of the Plant where the reference meteorological 'dispersion t factor) X/Q) s 1. x 10 7 sec/m 3'.
~
The total dose to t'e skin from noble gases is approximately 'equal the beta radiation dose to the skin plus'he gamma radiation dose to
'o the total body.
3.8 Pro ected 'Doses due to Gaseous Effluent Technical Specification 3.9.2.P)'.1 requires dos s due to radioactive s notarial released in gaseous effluent to be ~ro ected oue- a slue" er at
~ least once a month in'order to guide plant personnel in operating the:
radwaste systems. This requirement is satisfied by calculating doses according to equations for
- l. dose to air offsite due to noble gas gamma, radiation as in 0'3.k)
- 2. dose to air offsite due to noble gas beta radiation as in
- 3. 5, and
- 3. maximally exposed organ doses due to gaseous effluents+
othe". than noble gases as in $ 3.6.
~Radioactive ~~
in particulate form having
$ I-131) a.
half-life
'" +Y )+)V& )
'nd radioactive material greater than 8.0 days.'
~ . I ~ page 24 The radioactive releases are measured according to Technical Specification Table 3.9-3 and/or derived zs outlined 'n ODCH sections 3.4 and 3.6. The dose from gaseou's effluents is prospected by extrapolat'ng the dose commitment to date during the current quarter to inciud the entire quarter. T.fan analysis for a given radionuclide scned led Tl Table 3. 9-3
~ has not been made during the current quarter, the concentrat. on'easured~
in the most recently analyzed sample 's assumed to be representative of releases during the current quarter. X'3O) With these possible modifications in source term data, the beta. Dose to air is projected with
= 91 ..D)
Pp X where 91 = number of days in the quarter X = the number of days to date in the current quarter D = beta dose to
~ ~
air l (mrad) per g 3.5 The gamma dose to air is projected with P~= 91 .,D where D ='amma dose'to air (mrad) per $ 3.4. The projected thyroid dose ~ due to iodine and particulates in ga'seous effluents is calculated with the relation P . = 91 5 DH an X k where DH = the dose commitment to the infant,'s thyroid due to releases k represented by analysis k (mrem) Alternately, the projected personal dose to age group a and organ n, P , due to iodine and particulates in gaseous effluents may be calculated an'ith the relation P an
=91 D . C32) ank k
where D ankk
= the dose commitment to organ n of age group a of the maximally exposed person due to releases represented by- '.
analysis k (mrem) per S 3.6. l
I ~ ~ p>ge 4.0 'ose Commitmerit from Releases over Extended Time
- 4. 1 Releases durino 'a uarter Technical Specification 6.9.4.a.l requires an assessment of radiation doses arising from liquid and gaseous effluents from the plant during each calendar quarter. The assessment includes the following calculations of dose as desc'ribed by equations for 1; 'otal body and maximally exposed organ doses"due to liquid effluent via eating fish and shellfish liken from. the:.
cooling canals as in 5 2..4 . . *. - ,
~
- 2. total body dose due to noble gas Y as in 5 3.7.1 '.
skin dose due to noble gas P as in g 3.7.2 4 total body and maximally exposed organ doses due to gaseous effluents+ other than noble gases as in 5 3.6:
- 5. doses to air offsite due. to noble gas Y as in 53 4 and due to nob)e gas 0 as in 5 3.5.
The dose calculations are based on liquid and gaseous effluents from the Plant during each calendar quarter, determined in accord with Technical Specification Tables 3.9-1 and 3.9-3. Aqueous, concentration is est'mated"according top 2.2 on the basis of
\
quarterly averaged discharge flow. Quarterly averaged meteorological cond'- tions concurrent witn the quarterly ga'seous release acing evaluated are I used to estimate atmospheric'dispersion and deposition. The receptor of the'dose is described such that the dose to 'any reside near the Plant is unlikely to be underestimated. That is, the receptor
~ . ~
is selected on the basis of the combination of applicable pathways of exposure to'aseous effluent identified'n the annual land use census and maximum 'ground level X/Q at the residence. Conditions (i.e location, X/Q, and/or 'pathways) more conservative (i.e. expected to yield .higher calculated doses) than appropriate for the maximally exposed individual. may be assumed in the dose assessment. Seasonal appropriateness. of exposure pathways may also. be accounted for. F
~Radioactive ~~ I 131~
material in particulate form having a
+ i%l>M) half-life and radioactive greater .than 8.0 days.
1 page 26 Environmental pathway-to-dose tranfers factors used in the dose calculations are provided in Appendix A of this'D~i.
" 4.2 Releases durin 12 Months The regulation governing the maximum allowable dose or dose commitment to a member of the public from all uranium fuel cycle sources of radiation and radioactive material in the environment is stated n 40 C:"R Pare 190.
It requires that the dose or dose commitment to a member of the public C' from all sources not exceed 25 mremlyr to any organ or 75 mremlyr to the ~ thyroid. FueL cycle sources or nuclear power reactors ctner than the. Turkey Point Plant itself do not measurably or significantly increase the radio-activity concentration in the vicinity of the Pla..t; therefore, only radiation and radioactivity in the environment attr "curable to t'e Plant itself are considered in the assessment of compliance wf.th 40 CFR Part. 190. Evaluations of dose due to liquid and gaseous effluent required by
+
0 Technical Specification 3.9.2.g'o assess compliance are calculated as described by the equations for:
- 1. total body and maximally exposed organ doses due to liquid effluent via eating fish and shellfish taken from .the cooling canals as in/ 2.0
- 2. total 'body dose due to noble gas Y as in> 3. 7 1
- 3. skin dose due to'noble gas P as inp' 3.7 2 total body and maximally exposed organ doses due .to ga.seous ef'fluents+ other than noble gases as in 4 3.6. 'I The doses are calculated on the basis of liquid znd gaseous effluents from the Plant during 12 consecutive months, determined in accord with Technical Specification Tables 3.9-1 and 3.9-3. For the purpose of the Annual Radiological Environmental Report, doses are based upon releases during a calendar year.
+YE +K< ~ *Radioactive ~~I-131, ", -"" '~ and radioactive material in particulate form having a half-li e greater than 8.0 days.
page 27 Aqueous radioactive. material concentrations are estimated according w
.to g 2. 2 on the basis of annual averaged discharge flow Annual averaged meteorological conditions concurrent with annual gaseous releases being ~
evaluated are used to estimate atmospheric dispersion, deposition, and plume gamma exposure. Th'e receptor of. the dose is described. such that the dose.Co any, resident near the Plant is not 1'italy to be underestimated. The receptor is selected on the basis of the combination of app3.icable pathways of exposure to gaseous effluent identified in the arcual land use census and maximum ground level XIQ at the residence. Conditions more conservative. than appropriate for the maximally exposed person may oe assumed in the dose assessment. Environmenta) pathway-to-dose transfer factors used in the dose calculations appear in Appendix A. O.
Table 3-l Atmospheric Gase'ous Release Points at the Turkey Point. Units 3 and 4 Effluent Release Source Point Gas decay tanks Plan vent Radwaste Building Plant vent Auxiliary Building I Plant vent Containment Purge Plant vent No. 4 spent fue1 pit Plant vent No. 3 spent fuel pit Spent fuel pit vent Air ejectors Turbine deck Steam generator 31owdown vent blowdown
Table 3-2 Distribution of Radioactive Noble Gases in Gaseous Effluent from Turkey Point Units 3 & 4 Nuclide Release fraction a)b Ar-41 9. 2E-3 Kr-83m Kr-85m 2. 5E-4 Kr-85 2. 5E-4 ~ Kr -87 1.6E-4 Kr-88 2. 1E-4 Xe-131m 4.4E-4 Xe-133m l. 2E-3 Xe-133 O. 99 Xe-135m 8. OE-4 Xe-135 3. 4E-3 "Xe-137 Xe-138 3 7E-4 Based on measured discharge from Turkey Point Units 3 6 4 during 1978 thru 1980. To estimate radionuclide concentrations in a sample in which only the total activity concentration'has been measured, multiply the total activity concentration by the fraction of respective radionuclides listed here.
Table 3-3 Transfer Factors for Haximum Offsite Ai Dose Radionuclide Air Dose Transfe =ac"'mrs A Yi ~ mrad mr~Ad 3 3
>Ci sec/m 31Ci sec/m Kr-83m 6. 1E-7 9 lE-6 Kr-85m 3.9E-5 6. 2Z-5 . Kz-85 5. 4E-7 6 2E-5 'r-87 2. OE-l> 3. 3E-4 Kr-SS 4.8E-4 + Vh Kr-89 '.5--~'; Lq Kr-90 2E-.4. 2. 5E-4 Xe-131m '4.9E-.6 3.5E-5 Xe-'133m 1.0E-5 4. 7E-5 Xe-133 l. 3.E-.5 3.3E-5 Xe-135m .1. 1E-4 >E 5 Xe-135 . 6. 1E-5 7'.8E-5 '.
Xe-137 4. SE-5 OE-0 Xe-138 2.9E-4 1 5E-0
.. Ar-41 2. 9E-4 l. OE-4 I
Ref: Regulatory Guide 1. 109, Revision 1, Table B=.l ra L
Table 3-4 Transfer Factors for Haximum Dose to a Person Offsite due to Radioactive Noble Gases Radionuclide Dose Transfer Factors P). "p i. mrem mr em 3 3c Ci sec/m yCi sec/m Kr-83m 2. 4E -9 Kr-85m 3.7E-5 4. 6E-5 Kr-85 5. 1E-7 4 2E-5 Kr-87 1.9E-4 3. 1E-4 Kr-88 4.7E-4 7.5E-5 Kr -89 S. 3E -4 3; 2E-4 Kr-90 4.9E-4 -2. 3E-4 Xe-131m 2.9E-6 1.5E-S Xe-133m 8. Or.-6 3. 1E-5, Xe-133 9.3E-6 9.7E-6
,Xe-135m 9.9E-S 2.'3E-5 Xe-135 S.7E-5 5.9E-5 Xe-1'37 4.5E-S 3.9E-4 Xe-138 Z. 8E-4. l.3E-4 Ar-41 2.8E-4 8.5E-5 .
Ref: Regulatory Guide 1. 109, Revision 1, Table B-l.
~
~ g Table 3.5 REFERENCE HETEOROLOGY ANNUAL AVERAGE A'&iOSPHERIC DISPERSION FACTORS X sec Q 3 X/Q are annual averaged factors of, atmospheric dispe=sion of a mixed mode gaseous release from the Turk y groin" Plant Period of record: 01/01/76 to 1" /3 '77 BASE DIST@!~CE IH lilt c,S / KILO?'ETE< S hFTD DESIGN SECT DIST .ZS '
.75 1 50 2-SO 3. 5'0 4-50 50 7 00 Hl .40 1-21 2 41 4-02 . 5.63 7 24 B.e5 11 26 tile 0 8 9c.-07. } -9E'-07 8 'E-08 5 OE-0" 3. Oc.-GB 2 'E-03 1.9E-OS 1-4E-OB N
p)C 0 '
~
6;9E-07 1.SE-07 6.3E-QG 7 rr CB 1" -08 1 3E-08 OE<<nS EtiE 0. 8.4c-07 1.4rE-07 Sc GO 3. 9E-GG 2.6":-QG 2 '=--08 1 ' -09 1-v=-08 E B.GE-07 'i 9c 07 9.}E-08 S-}E-08
~ 3 6 n8 2.7E-08'.=E-OB 2.2E-GB }.7=-00 ESE n. 6.6E-07 )-SE-07 7.9E-GB USE-00 1.9=-08 }-2E-Oil SF 0 }.6E-o6 7..BE-o7 }E-07 6-}E"00 4.2E-CB 0" -08 .2.6E-08 2. 1 c.-08 sE 0 4.9E-C5 9-2E-07 3.6E-07 }.BE-07 1. }E-07 9.0c-08 7.1c-08, -08 S 0 ~ '2-9E-06 6c p7 1.8E-07 1-PE<<07 7 Qc 0<} 5.4=-08 '4.5E-OS 3.3E-.OS SSM 6-SE-07 }.6E-07 6.5c-08 4.6c-08 Z, ~ E-eo 2.5E-OB 1.8=-'OB }--'..-OG SM 0 3.?c-07 }.4"-07 7.9r-08 4 'r"GB 3.2E-08 2.7E-08 }.9E-CS Msd 2.9E-G5 5 3c,-07 2.3E<<07 1.3" -07
~ 7a6E-08 5.'5" -OB 3.1c no Oi 6.3E-05 } 3E-05 SIZE-{)7
~ 6c 07 }.7E-07 2 07 9.2=-08 6E-0 8.}E-08 6.3E-n3 4-}E-06 8-7E-i'7 1 ~ 7E-07 Zc.-.07 0
0
~
7.7E-P6 6.0E-'0. 2 4" <<07 1. 2E-07
~
7'c n3 5.}E-OB 4.3E-OS Zc 0{.. CB G }-Ac-06 2-9E-07 }-ZE-07 5-85-08 4.5r-CB OQ 2.4c-p8 }-SE"CD 0 ~: 9 '=-07 2 'E-07 8. c-O8 4 5 -08 3.7E-CB 2.2E-OB 1.7c.-OB 13-0 BASE, OISTAWCE'IH ((ILES / KILO"'<ETERS AFTD OF >I G'rP DI ST 1}.00 .79 5.00 }.00 2 ~ 00 2.75 30 SECT HI GO
}4 '40 17.70.. ,}.27 8 '4 1 61 3 ~ 22, 4-42 4
5.92
. ONE Oo 9.$ E-09 6-6E-09 }.GE-07 2-pc-0 1 ~ 4E-07 6-2=-08 4.4E 2-3E-08 t4c, 0 7-3c.-09 5.4E-09 1.5c-"07 }.GE-08 }. },E-07 4. ue.-OS 3.5c,"CG Ei~E .0 1.1E-08 7-4E.-09 1. 4" -07 2.0E-OB 1 ~ OE-C7 5 ZE-OB 6c 4E Q 0 }-3E-08 9 8= P9 1 7c-07 }.3c-07 6.==c-ps 4 <r 08 2.- C<E-0 ESE 0 }.ccIE-oB 9 5E-C9 }.4c-0'7 <7 08 }.ZE-O? 5 7E-OS -.' OE-n S SE 0 ~ 1 ~ 3c.-CB ?.7E-07 2 '=-08 1, 9" -07 7. GE-08 S .:-=-0 9 -"- ~ c.-p ssE 0 ~ 3~<E-08 2,7E-OR 8."c-07 7 9E 08 6..;E-07 2. 5E-07 }.6=-07 4c n S 0 ~ Zinc 08 }.SE-CB 4'.25-07 S.pc-GB 3. }r-07 }.3:=-07 9.5E>>nB SSM .0 ~ 9.4E<<09 7.}E-09 }.SE-07 'c-GB 2 1,1E-C7 5.4c-08 0 5 1.{ n. }.4E-OG }aCE 08 nc 07 7.9c--PB ?c 07 ~-QC-07 6.9E-'08
'4 5'n 0. 2-2"-08 }-BE-08 5.9E<<07 4.8" <<GB 4. iE-07 1.7E-O7 Oc-n7 Oi 4,, E-08 3.SE-QB '-?c-06 }.OE-07 9 OE-07 :c-07 } ""c-0 5.9'='"07 2. 3'07 1 . GE-07 8 io~-0
<i<
'1<'O'J C. 7.9c-08 2--E-08 8 '-07 7. }>>-08 tip 0. 2-GE-OB }.SE-08 5 'r-07
~
c-0:i 4.}c-07 1-GE-07 4
~ .'
VV<<I 0 }.GE-08 09 7,7E-07 2.6E"08 7 i~c-'
~ 9 }c Og 6 i c-p.
M 0. prp8 -...".=--09 } 9E-07 ~ pc-0$ } . 5:--07 5 9c-{i8 CE-O& Za3< 0 {ai.~:-:Eo OF'."-LID OS"-ERVI T 1045 }6~38 {jUut7go GF }'WV<LIO (l3 >E<"-;VAT}0<{5 loo6 195 V>~<V.P. r7 C'LHS V~PEn. LEVEL 383
~ ~
Table 3.6 REFERENCE HETEOROMGY DEPOSITIOii< DEPLETED ANNUAL A<<)ERAGE A~l<1OSPHERIC DISPERSION FACTORS t X sec Q X /Q are'annual averaged factors of atmospheric dispersion of a d mazed mode release from the Turkey Point Plant which have been corrected for depletion from the plume by fallout and deposition. Period of Record: Ol/01/76 to 12/3I/77 BASE DISTANCE IN l>><ILES / KIl OWETEPS rFTO DESIGN SECT DlST 25 '75 1 50 2.50 . 3-50 . 4-50 5 50 7.00 HI -40 1-21 2.41 4 02 5.63 7-24. 8 05 11-26
.t(NE 8.7E-o7 1.7E-o7 7~3E-OG 4 ~ 4E~O8 2 i 7"-08 1 9E-OFi 1-2=-00 t4E 0 6.9E>>O7 1-4E-07 S.SE-.OG 3.3E-oe 7.2" -08 1 ?E-GG 0 Pl 8 Gc-09 ENE 0 8.0c-07 1-2E-07 e.sc-OG c OG 2.4E-OB 2 Oc 0 1.65-08 E 0 8.6:--07 1.7E.-07 7.6E-OG 4:-4E-0<) 3.1E-CG 2 4c-0~ 1.9"-cG E.S=-OO ESc. 0 6.1E-07 1 3" -07 6-9E-OG
~ 3.9=-OB 2.'Sc-OG 2- GE-OG 1 6c-0 3 SE G 1 c 2.6~-07 ,9.5E-OR 5-2E-OB 3.4=-08 2- -'. E-08 2 1E-OB 1 e7c. OG ssE 0 ~
-06'-7E-05 G.2E-G7 3 lc 07 5 -07 9.2E-GG 7 4E-08 5.8E-OB 3. 8" -08 S 0 ~ 2.ec-06 4-2E-o7 1.SE-07 0-cr-00 6.4E-OG C 4- -GG 3.7E-O8 2-.6c.-ce HS4'i 5 '5-07 3. 9"-OG 2.0E-GG ?-2c-08 t
SSM G.'. 4E-07 5-5E-08 SM 1- .E-G6 2.8L-07 l ~3c. 07 6 7:--OB 4.2< -08 2-c -08
~ 2 3c.-'3 =
G. 2.75-05 5-6E.-07 2. 1E-07 1 . Gc.-07 6.4E-O8 .<E-0<s 3 Sr 0<8 2.:-E-GP. 0 ~ 5 9>> 05 1-2c-06 4.4E-07
~ ~ c.'cE- 07 1 ~ 4F 07 9.9" -<3~ -08 5 ~ 4" -08 MHH 0, '3.'ec-06 7.7E-O7 2.9E-07 9.8E-OG 7. OF-08 5 sc. Gc<t% 3 6=-OP N'8 0 2.5 -0 5.42-07 2,1=-07 l.iE.-07 6.8E-OB 4-SE-03 3.8E-G8 2-GC-OR NN'rl 0 ~ 1-4=-G5 7.. 6c.-07 1.1C-O) Or 08 4.0E OP 2 'E-,GG ""-08 H 0. G ~ Gc.-07 3>> 9E-07 7.Fsc-OB 3.9E-OG 2 'E-CG 1 9.":-0 8 1 . SE-08 1-'E-OQ BASE OIS T:1hcE EN )-: ILES l LO),ETr RS AFTO DES)GM sECT DIST 9.00 11 ~ 00. .79 5 00>>
~ l. GO 2 QO 2-75 ~ 4 ~ 30
).( I 14.48 17 70 1.27 8.04
- 1.61 3-22 6.92
'N E 0 ~ 8-5E-09 6-0" -09 1 6E-07 I ~ GE-OB 1.?E-.07 5.SE-OG 3e vl OG 2.,1C-OG NE C ~ 6-3c-09 4- 5c.'-09 1 ~ 3E-07 !.4E-08 9 4E 08 2" -08 nc -Oon ENE Os 9-0E-09 6.7E-09 1 - 2E-07. 1.SE-OG 9. 1E" 08 4.5"=-oG q E-08 E 0 1-1E-GG 7 oc Go 1 ~ SE-07 2.1EroG 1-PE-07 S.S:--oe 9c-08 2&4 Oc; 8-BE-09 8.3c-09. 1 . 3c-07 ) .c'"-08 1 OE-07 5 GE cp~ 3.4c-OG 2 CC-0=
'sE C~
SE, oi 1;3E-OG 1.0C-GG 4--Og 2.3:-08 1.7E"07 7c 4.7=-08 ssE 0. 7.7E-OG 2 '"..-08 7'-7E- 07 6.4E-OB 5. 6.=.-07 2 ~2c- 7 5 0 1-9c.-oB 1 3E-G8 ~ 3. Gc-07 4.>c-OG 'E"C7 1.1E-07 7. l;E-O.) 4 gC-0 ~: 2 SS l) 0 7 9c 09 5.7c-09 '.4c-07 1.GE-OR 9.6E-CG 4.7C-08 3 2'=-OR 2-2" -Cr S'i 0 ~ l.lE-O>>8 G 6" -09 ~ 2.7E . 07 2-'c-08 2. Oc.-07 9.1E-OG 2- ~c.-o MSH 1.GE-OG 1.4E-OG S-2E-07 (,. 0<c OB 3-GE-07 1 . 4E-07 '
'.82-0:
- 7) r,. 3.7E-oG 2 Gc-0 ~ 1-lE-06 >>s. c" -Oc c-07 3 1 c-07 2. 0"-07 0
- 0. 'E-09 7 'E-07 5e fE OG )E 07 2.0c-07 .} . 4E-07 7 4 4" n
')
ht <<) ? USE OG 2 ~ ~ NW Oe 1.8E-OG E 08 5. 1E-07 4 'E"OS 3.6E-07 1 ~ 4c;-0,7 5 ~ 0< 0
) Wn' e. 9.lr-09 ~
oc'.3E-09 7.4C-O7 2.3E-GR 1.-".E-C7 7 r.. O<s 4E-GG
- 0. 8 7E-09 1 ~ eE-07 1 ~ 7C-08 5.ZE-CG 3 ~ SE-OG
) <l)'(Gc iR Oc VA<. EO OBSE~<.<A-EOWS 16S3G MU) <3E~ OF 'H<<,'~'0 l. -Rv> I <-'S < lo05 Nl)H P. OF CA> 'sS LO '=e 195 N<lg!S..:".P. OF ( <?pep LF</EL 383
~ >
~ ~
I> Table 3,7 REc'EREHCE HETEOROLOGY ANNUAL AVERAGED RELATIVE DEPOSITION RATE D 1 Q D/Q are annual averaged factors representing the ractxon of a maxed mode airborne release from the Turkey 'Point Plant which is deposited on a square meter t area o f land at a given distance and direction. from tne of Record: 01/Ol/76 to 12/31/77 Plant.'eriod BASE DISTANCE IN nlILES l KILCRETERS
'FTO" DESI G>N SECT DIST .25 .75 )-50 F 50 .
3 50 4i50 5-50 7.00 HI ~ 40 - 1 21 2 41 4 02 63 7-24 0 GS 11-26 Q ~ 6~4E-09 1 . Sc.-09 4<7E-10 2- GE--I 5 SE-}I 4 )E-'1 2.7E-) 1 t'E 0 3~5E-09 8 7"-3 0 2 8E-}0 ) -2~c-}i> 4E~)l Sr 7> ENE n.. 2.8E-09 S.lE-}0 2.}E-10 7 ec.-} 1 } )1 2.9":-11 }.9E-1 l E 0 '.7E-O9 6.6E-)G 2 'E-)0 I 1""" 10
~ 8r-I 1 3 '~"}} 2 rr-)1 ) .6F-})
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).2E-08 2.)E-09 6c7E-3 0 3-0E-)O 2 0 -)0
-10'}E S 0 2 10 <1 SS>". 0 ~ 2.3C-09 7-?c-)0 2.4E-)0 1.2E-)G 5 'E-)) 4-i">E-} l. 8r 4, SM Oo ).)E-GG 2.7c.-09 ) -0. -09 4 3E ) Q
~ 2 'E-)0 }.2E-)0 9 6c-} ). 5
'8SQ 0 2-3c.-08 5,0 l. SC-09 6 }E )0 ~ ?C; }Q 2-GC-10 1 4E-I 0 ~
O. 5.7E-08 1.2E-Q8 3 c c 1 . 4E-09 7 6>>-10 4.9E-)n 3. 3E-}0
'4'NU 0 4.<C-08 %.6F-09 2a7-"09 1.0c-"09 5 7c }0 3 'E-)0 2.4E-)O
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~ 9.6=-)1 5.8!:-}. 4 G=-l 1 2.5E i I ~
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le 27 8 '4 1 61 22..4 6.92 NME 0 ).6c.-)) 9.'3r -32 } 4E-09 4 e 9~6E-)0 ?.8E-)0 ).6E-lG 6-2c.-3.1 YE.. 0 ~ 9 9>>-)2 6 '2E" 12 ri.}c-)0 7E-1'.2E-} 1 5. 6E-1 0 8E 1 0 '1 C 1 ~~:l Gr-< 1 EWE G ~ ~ 8,)E-12 S. 2=-}2 5,0E- 0 '.3E-}l 3.6E-}0 1.2E-)n 6.4E-) 1 3. 0=-} 1 c 0 }rOE-)) 6.6c-)7 5<<9E-}0 3-pc-')l 4.3r-)0 )-SE-}0 8 pl 3.9=-11 CSE 7.5 5.8E-)2 4 ) 10 2.2=-)) 3 1""10 ).2>>-10 Sc 8~-} 1 SE 0 0 ~ )-3~-)) '"-G9 4 1 '=-)l 7 1 -)0 2-3=-)'0 ) 3c.- I 0 6 0
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APPENDIX A. PATHMAY-DOSE TRANSFER FACTORS Environmerital pathway transfer factors, usage factors, and dose commitment factors appropriate for each exposure pathway> age> and organ are combi'ned into integra.ted environmental concentration-to-dose, factors for each radionuclide. This appendix includes tables of values of the transfer factors calculated in accord with equations and values 1 recommended in NUREG-0133 for individual environmental pathways. Tn the event a single, composite transfer factor is desired for a given or'gan and age group, it can be obtained by summing the factor's'or appropriate pa thways. Appropr iate transfer factors from Appendix A are used in performing dose assessment calculations prescribed in the ODCH. J. Boeglly et al., eds., 1978> Pre aration of Radiological Effluent Technical Soecif ications for Nuclear Power Plants, NUREG-0133, USNRC, Office Nuclear Reactor Regulation.
Appendix A Tables are the same as in the August 14, 1980 version
APPENDIX B Nuclide Dilution in the Coolin Canals The effective concentration of each nuclide in the cooling cane.ls is a function of the discharge flow and concentration, the nuc'ide decay constant, and the various diluting flows in and out of the canal Reference.l was a study of the hydraulics of the Turkey Point. cooling . canals it revealed that ground water interchange is the most significant means by which net removal of chemical constituents from the canals occurs. One result of the study documented in Reference 1 is a mathematical model of the water and chemical balance in the canals. That model is used here to derive an expression for the nuclide concentrations to which fish and shellfish in the canal are exposed. This expression can then be used to develop the expression in Section 2.4 for calculating accumulated dose due to ingestion of fish and shellfish. 'ersonal Assuming that sampling of nuclide i in the canal by seafood (and, in turn, the ultimate receptors) occurs randomly in space and time, the canal can be treated as a homogeneous mixture. This is certainly true for nuclides with effective lives greater than several days, since the canal average 'transit time is on the order of two days. For shorter-lived nuclides, this 'assumption is conservative. (See also Figure 12. of Reference 2.) A mass balance on the canal yields: V dC. i {t)/dt = i i (t)
-V) .C. - F 3 i (t)
C. + F C. 1 ik ~ (al) where .V is the canal volume (ft ) 3
'is the nuclide i decay constant (min -1 )
i C. i (t) =. is the average nuclide i concentration in the canal't time t (pc/ml)> F is the ground water interchange flow ( ft 3 /man) flow in radioactive liquid discharge line (ft = 3 /min) radionuclide i concentration in.undiluted liquid waste being discharged (PCi/ml)
~ ~ Defining C.(t) = C. when t = o at the beginning of the discharge, it can be shown that: i, i C..(t) where the steady-state
= C. - (C. - C.) exp i
concentration I - (X. + F 3
/V) (B2) i, C. = P i
C. = P 1 1 C. ik/(X.V+ C.k/Vg. ik i P i
) or (B3)
(B4) Examination of the above equations reveals that the max.mum possible concentiation of nuclide i is C.. s i For continous discnarges, C s is i approximately equal to the "average" concentration in the cooling cana.l. For batch discharges this expression is conservative, part,icula=ly when the duration of the discharge is small with respect to the period between e discharges and the effective decay period, I/X.. Pigure 27 of Reference 1 show's the relat'ionship o'f canal water level (Canal 32 - Line 3) to volume. Figure 9 of Re erence 1 shows . canal levels during July - December 1978. During tnat period, tne canal volume (V) never decreased below 500 million cubic feet, - a conservative lower bound for V. Tables 3 and 4 of Reference '1 provide an expression for F , the ground water interchange flow rate
. F 3
= C 1
+ C 2
+ C 3
+C' (ft3 /min) (B5) where C 1
= 30 900 ft /min. '
C = + '1680 neap/spring C = -.120 June 3
. +.'360 J.uly
+ 1320 August
+ 1200 September
-,960 October
- 1440 November C = 3000 A conservative lower bound for F 3
would be on the order of 30000 cubic feet. per minute.
An upper bound on the steady-state concentration of radionuclide i in the canals is thus s s s C. Q F . C. F . C.~ (B6) i 3 5 x 10 ft .- X. + 30000 ft /min Alternately, stating Fl and F3 in units, gal/min, and V in units, gal> produces or s C.Q F 1 V+F'95
. C.~
k 3 3.75 x 10 1 Fl gal C.~ i+
)%.. 2.24 x 10 ga1/min i 3 3.75- x 10 {X. + 6-. x 10 )
s or C.( F . C. V'.
= 9
+here V 3:75 x 10 gal e -5 . -1
+ 6 x 10 min
. i ,
References
- l. Evaluation of Coolin S stem Chemistry, Turke Point Cooline Canal
~Sstem, Dames and Hoore, June 1979.
2.:stima tin Aquatic Dispersion of'ffluents from Accidental and Routine Reactor Releases for the Purpose of Xmolemen'tin Append jx X., Regulatory Guide 1.113, Revision 1, USiXRC, April 1977.
N ~ ~ \ ~ APPENDIX C Limited Analysis Dose Assessment for Liquid Radioactive Effluents The radioactive liquid effluents for the years 1978, 1979, and 1980 were evaluated to determine the dose contribution of tne radionuclide distribution. This analysis was performed to evaluate the use of a limited dose analysis for determining environmental doses. Limiting
-the dose calculation to a few selected radionuclides that contribute the majority of the dose provides a simplified method of determining compliance with the dose limits of Tecnnica) Specification 3.9.1.b.2'.-
'Tables C-l and C-2 present the results of this evaluation.
Table C-1 presents the fraction of tne adult total body dose contributed
'y the major radionuclides. Table C-2 presents the sar e data for the adult GI-LLI dose. The adult total body and adult GI-LLI were determined to be the limiting doses based'n an evaluation of all age groups (adult> teenager'> child, and infant) and all organs:(bore,,
liver> kidney> lung> and GI-LLI). As the data in the tables show, the radionuclides Co-58, Co-60'> Cs-134> and Cs-137 dominate the total body dose; the radionuclides', Co-58, Co-60> Hb-95> and Ag-110m dominate the GI-LLI dose. In all but one case (1978 shellfish, GI-LLI-dose) these radionuclides contribute 907.'r more of the total dose. If for 1978 the fish and shellfish pathways are combined as is done to determine the total dose, the contribution from these nuclides is 897. of the total GI-LLI dose.
~
Therefore> the dose commitment due to radioactive material in ..; . liquid effluents can be reasonably estimated by limiting the dose calculation to the radionuclides, Co-58, Co-60> .Hb-95> Ag-llOm> Cs-134> and Cs-137, which cumulatively contribute about 90/ or greater of the total dose calculated by using all radionuclides detected. Th's lim'ted analysis dose assessment method is a simplified calculat'on that p"ovides z reasonable evaluation of doses due to liquid radioactive ef'=luents.
~ 0 ~ g J Tritium is not. included in the limited analysis dose assessment for liquid releases because the potential dose resulting from norma 1 reactor releases is negligible: and is essentially independent of radwaste 'system oper'ation. 'he maximum amount of tritium released annually is about 1000 curies. At Turkey Point> 1000:Ci/yz releases to the cooling canals produces a calculated whole body dose of 0.002 mrem/yr via 'the fish and shellfish pathways. This amounts to less than 0.17. of the design objective dose of 3 mrem/yr. Furthermore> the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste'system operation.
~ ~ I 1 e Table C-1 Adult Total Body Dose Contributions Fraction of Total (Excluding H-3)
'Radionuc1ide 1978 1979 1980 Pish Shellfish Fish Shellfish Pish Shellfish
~ Co-58 0. 02 0. 12 0. 02 0. 10 0.43 0,4)
Co-60 0. 11 0. 57 0. 12 0.56 0. 23 0.68 Cs-134 0. 41 0. 13 0. 43 0. 13 0. 27 0. 05 Cs-137 0.44 . 0.14 0.40 0. 12 0. 43 0.08 Total 0. 98 0. 96 0.97 0. 91 0. 96 0. 90 Table C-2 Adult GI-LLI Dose Contribution Fraction of Total (Excluding H-3) Radionuclide 1978 1979 1980 Shellfish Fish Shellfish Fish Shellfish
'ish Co-58 0. 07 0.15 0.05 0. 13 0. 03 0. 10 Co-60 0.33 0.70 0. 27 0.67 0. 14 0.69 Nb-95 0.54 ( 0.01 0.13 4 0.01 0. 56 (0. 01
. Ag-110m 0. 48 0.12 0. 24 0. 12 Total 0.94 0. 85 0.93 0. 92 0.97 0.91
APPENDlX D Technical Bases for A e ff Overview The evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfe factors instead of using dose factors which are radionuclide speci~ac. TheNe effective factors, which are based on the typical radionuclide distribu'tion in the releases, can be applied to the total radioactivity released to approximate the dose in the environment, ie, instead of having to sum the isotopic distribution multiplied. by the isotope specific dose factor only a single multiplication (A ~ times'he total quantity of radioactive material released) would be needed. This aoproach provides a reasonable estimate of the actual dose while eliminating the need for a detailed calculational technique. Determination of A ei~f The effective dose transfer factor is based on'ast operating"data. The radioactive effluent distribution for the past years can be used to derive a single effective factor by the following equation.
~ A..
W i f.i A = eff where A ff eff
= the effective dose transfer factor A,=the dose transfer factor for radionuclide i f.i = the fractional abundance of radionuclide i in the radioactive effluents T Th's equation yields a single dose factor, weighted by the typica1 radionuclide distribution.
1 ~ I g~ ~
\
To determine the appropriate effective factor to be used and to evaluat~ the degree of vari'ability, the atmospheric radioactiv'e ez )uents zor z h pastt 3 yearss have the a e been evaluated. An effective dose transfer factor has been determined for the gaseous effluents for all pathways of interest. Tables D-1 and D-2 present the results of this evaluation For the radioiodines,and particulates with half-lives greate- than 8 days, the effective dose transfer factor is based solely on the radioiodines (X-131, 133, and 135). This approach was selected because the radioiodines contribute essentially all of the dose to the infant's thyroid yia the cow-milk pathway. The infant's thyroid and the cow-milk pathway are the critical organ and controlling pathway, respectively, for the releases of radioiodine and particulates. All other particulates contribute less than 1/ of the dose. The effective dose transfer factor is determined by applying equation D-1 to the radioiodines. However,'indetermining the dose, this effective dose transfer factor should be applied to the total release of all radioiodines and to particulates with half lives greater then 8 days.,This uniform application is conservative in providing reasonable assurance that the actual dose will not be unde-.estimated by the use of this simplified method. Th dete~mination of Aeff was limited to t'ne past three years
~
(1978 1979 an and 1980) because, of the changes tha" have occurred in the waste processing system. A demineralizer system replaced the previously
'used e'vaporator in the liquid waste processing system.
As dan be seen from Tables D-1 and D-2> the efzective dose transfer' factor varies 1 ittl e fr o m y year to year. Th'e maximum observed va iabilxty zrom the average value is 13K for the noble gases and 2S7. for the radioiodines. This variability is minor considering other areas of r uncertainty an d consservatism inherent in the environmental dose calculational models.
i To provide an additional degree of conservatism> a factor of 0.8 is i'ntroduced into the dose calculational process when the effe'ctive dose transfer factor is used. This added conservatism provides additional assurance that the evaluation of doses by the use of a single effective factor will not significantly underestimate any actual doses in the
~
environment. Reevaluation The. doses due to the gaseous effluents are evaluated by the more detailed calculational methods (ie, use of nuclide specific dose factors) on a yearly bases. At this time a comparison can be made between the simplified method and the detailed method to assure the overall reasonableness of this limited analysis approach. - If this comparison
'I t
indicates that the radionuclide distribution has changed significantly c'ausing the simplified method to underestimate tne doses by more than 207.> the value of A ff will need to be reexamined to assure the overall eff
. 'acceptability of this approach. However> this reexamination will only be needed if the doses as calculated by the detailed analysis. exceed 50K of the design bases doses (ie, greeter than 5 mrads gamma air dose, 10 mrads beta air dose, or. 7.5 mrem infant thyroid dose).
In any case> the appropriateness of the Aeff value will be periodically evaluated to assume the applicability of a single effective dose factor for ev'aluating environmental doses.
e 0
I e~ < ~
~ ~
Table D-1 Effective Dose Transfer Factors Noble Gases Air Dose A A8 Year ff eff IQI ad mrad pCi . sec/m 3 Ci . sec/m 3 1978 1.3 K 10 3.4 x 10
-5 1979 1.3 x 10 3.4 x 10 19,80 1;6 x 10 3.4 x 10 Average 1.4' 10 3.4 x 10 e
Table D-2 Effective Dose Transfer Factor for Air-Grass-
'Cow-Milk-Infant-Th roid Pathwa Radionuclide Annual Fraction Dose Weighted Airborne Factor & Dose.
Release mrem/ r Factor (Ci) gCi/(m 2
. sec) mrem/ r ~ g VC /(m . sec)
Year 1978 I-131 0. 381 0.688 9.9E11 I-133 0. 129 0.233 1. 3E10 6. 9E11 I-135 0.044 0.079 5. 2E6 Year.1979 I-131 0. 0188 0.520 9.9E11 I-133 O'. 0156 0. 432 l. 3E10 5. 2E11 I-135 0.0018 0. 0848 5.'2E6 Year 1980 e I-131 I'-133 I-135
- 0. 0518
- 0. 0124 0.0043
- 0. 756
- 0. 181
- 0. 063 9.9Ell
- 1. 3E10
- 5. 2E6 b
7.5Ell avg 6. SE11 air-grass-cow-milk- infant- thyroid dose transfer fac tor b Effective dose commitment transfer factor is the average of weighted dose transfer factor over three years.
APPEHDEX F Bases of Technical Specification 3.9.l..d.l Limits I Technical Specification 3.9.1.d.l conditions for treating radio-active liquid waste are stated on a quarterly bas's as one-fourth of the design objective (annual) limits on individual dose commitment. At the Turke'y Point Plant> the individual dose limits require lower releases of radioactive material in liquid effluents than would cost-benefit cri.teria for treatment. This occurs because the closed cooling canal system yields a'ow population dose. Consequently> a liquid radwaste system operability requirement can be based on Appendix '1 design basis limits without any further reduction for treatment cost-benef icial ity. I ODCM sections 2.4 and 2.5 present the dose projection method used to decide operability. Since 90 percent or-more of the dose is normally caused by' limited number of radionucl'des, dose projections can be adequately based upon those radionuclides. Appendix C explains the basis for this position.
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TURKEY POINT PLANT REFERENCE TO nAP LOCATIONS DES I GHAT ED DESIGNATED LOCATION .NUNBER LOCATION NUMBER T51 1 H-1 24 T52 2 N-5 25 T56 3 T57 4 HHW-1 26 T58 5 HNW-10 27 T63 6 NW/WNW-1 28 T66 7 NW-5 29 T71 8 NW-10 30 T?2 9 W/ma<-5 31 T73 10 WNW-10 32 T74 11 W-1 33
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