ML20140D334

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Rev 2 to Offsite Dose Calculation Manual. W/Two Oversize Maps
ML20140D334
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 02/28/1986
From: Leach E, Perkins T, Roman T
NIAGARA MOHAWK POWER CORP.
To:
Shared Package
ML17058A533 List:
References
PROC-860228, NUDOCS 8603260064
Download: ML20140D334 (73)
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Text

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _

NINE MILE POINT NUCLEAR STATION l

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NINE MILE POINT UNIT 1 0FF-SITE DOSE CALCULATION MANUAL (ODCM)

{

DATE AND INITIALS APPROVALS SIGNATURES REVISION 2 REVISION 3 REVISION 4 Chemistry and Radiation Management Superintendent A//cr'p(

E. W. Leach 86) -

M (M Station Superintend NMPNS Unit 1 T. W. Roman . ' ugh _ ( Ak General Superintendent

/( /' ' /

Nuclear Generation g gp,,

T. J. Perkins' Is I . [A bh If

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d Summary of Pages Revision 2 (Effective )

. NIAGARA MOHAWK POWER CORPORATION T$IS PROCEDURE NOT TO BE USED AFTER SURIECT TO PERIODIC REVIEW.

i 8603260064 860228 PDR ADOCK 05000220 R PM

OFFSITE DOSE CALCULATION MANUAL NINE MILE POINT UNIT 1 i

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February,1986 I -

ODCM - NINE MILE POINT UNIT 1

1.0 INTRODUCTION

2.0 LIQUID EFFLUEh7S 2.1 Setpoint Determinations 2.2 Dose Determinations 3.0 GASEOUS EFFLUENTS 3.1 Setpoint Dsterminations 3.2 Dose and Dose Rate Determinations 3.3 Critical Receptors 4.0 40 CFR 190 REQUIREMENTS . y 4.1 Evaluation of Doses From Liquid Effluents 4.2 Evaluation of Doses From Gaseous Effluents 4.3 Evaluation of Doses From Direct Radiation 5.0 ENVIRONMENTAL MONITORING PROGRAM 5.1 Sampling Stations 5.2 Interlaboratory Comparison Program 5.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements TABLES <

APPENDIX A- Dose Parameters for Iodine - 131 and 133, Particulates and Tritium B- Diagrams of Liquids and Gaseous Radwaste Treatments Systems C- Dispersion Calculation Tables

1.0 INTRODUCTION

The Offsite Dose Calculation Manual (ODCM) provides the methodology to be used for demonstrating compliance with the Radiological Effluent Technical Specifications (RETS),10 CFR 20,10 CFR 50, and 40 CFR 170. The contents of the ODCM are based on Draf t NUREG-0472, Revision 3. " Standard Radiological Effluent Technical Specifications for Pressurized Water Reactors," September 1982; Draf t NUREG-0473, Revision 2, " Radiological Effluent Technical Specifications for BWR's," July 1979; NUREG 0133,

" Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978; the several Regulatory Guides referenced in these documents; and, communication with the NRC staff.

i Section 5 contains a detailed description of the Radiological Environmental Monitoring (REM) sampling locations.

Should it be necessary to revise the ODCM, these revisions will be made in accordance with Technical Specifications. .

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2.0 LIQUID EFFLUENTS

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2.1 Setpoint Determinations s

4 2.1.1 Basis Monitor setpoints will be established such that the concentration of radionuclides in the liquid affluent releases in the discharge canal will not exceed those concentrations as specified in 10 CFR 20, Appendix B Table II, Column 2. '

Setpoints for the Service Water System Effluent Line will be calculated quarterly based on the radionuclides identified -

during the previous year's releases from the liquid radwasta 4

system or the isotopes identified in the most recent radwaste release or identified probable source. Setpoidts for the Liquid Radwaste Effluent Line fil be based on the radionuclides ,

identified in each batch of liquid waste prior to its release.

1 After release, the Liquid Radweste monitor setpoint may remain t

as set, or revert back to a setpoint based on a previ'ous Semi-Annual Radioactive Effluent Kalease Report, or in; tall blank flange in the discharge line and declare inoperable in accordance with the technical specification.

2.1.2 Service Water System Effluent Line Alarm Setpoint The detailed methods for establishing setpoints for the Service Water System Effluent Line Monitor shall be contained in the Nine Mile Point Station Procedures. 'Ihese methods shall be in accordance with the following:

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.--.,,--y-- -e,-- , , ewe- .- - --,--- - . - , -a W - .y ---,-v-w -- -- ,- -- - ,-s- , - - ,, , , - - - - - , -- - + - - . . , . -

Setpoint (Hi-Hi alarm)<0.9 (C1 ) TDF CF + background F,, Si .

MPCi where C y = concentration of isotope i in the sample (units = uCi/al)

TDF = Total Dilution Flow (units = gallons / min)

F,, = Service Water Flow (units = gallons / min)

CF = monitor calibration factor (units = net cps /uCi/al)

MPC =

f liquid effluent radioactivity concentration limit for radionuclide i as specified in 10 CFR 20, Appendix B, Table II, Column 2, for those nuclides present in the previous batch release from the liquif radwaste effluent system or, r for those nuclides present in the last Semi-annual Radioactive Effluent Release Report (units = uCi/al) or for those nuclides present in the service water system.**

0.9 = factor to account for the presence of pure beta emitters i

(= Required Dilution Factor HPC g TDF = Actual Dilution Factor F3y

    • For periods with known reactor water to RCLC system leakage, RCLC maximum permissible concentration may be prudently substituted for the above.
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Setpoint (Hi alarm)<0.7 (Cg) TDF CF + background F,, $1 MP Cs -

2.1.3 Liquid Radwaste Effluent Line Alarm Setpoint The detailed methods for establishing setpoints for the Liquid Radwaste Effluent Line Monitor shall be contained in the Nine Mile Point Station Procedures. These methods shall be in accordance with the following:

Setpoint (Hi-Hi alarmh<0.9(IC1 ) TDF CF + background hcg where C g =

activity of isotope i in the sample (units =

uCi/ml).

TDF = Total Dilution Flow (units = gallons / min)

F, , = Radwaste Effluent. Flow (units = gallons / min)

CF = monitor calibration factor (units = net eps/uci/al)

MPC g =

liquid effluent radioactivity concentration limit for l

radionuclide i as specified in 10 CFR 20; Appendix B, Table II, Column 2, for those nuclides detected by l

l spectral analysis of the contents of the radwaste tanks to be released (units = uCi/al) l l

l 0.9 = factor to account for pure , beta emitters l

C d-

=

Required Dilution Factor MPC g 4_

g._..

_. . 2.- ; -- - _ _.__ _ _ _ . _ . _ _ , _ _ . _ . _ .

l .

Setpoint (Hi alarm) = <0.7 (C1 ) TDF CF + background Fra Ei MP Ci Note: ir TDF = gC F mci the discharge could not be made, since the monitor

.would be continuously in alarm. To avoid this situation, F,, will be reduced (normally by a factor of 2) to allow setting the alarm point at a concentration higher than tank concentration. This will also result in a discharge canal concentration at approximately 50% maximum permissible concentration.

2.2 Dose Determination 2.2.1 Maximum Dose Equivalent Pathway A dose assessment report was prepared for the Nine Mile Point Unit 1 facility by Charles T. Main, Inc., of Boston, MA. This report presented the calculated dose equivalent rates to individuals as well as the population within a 50-mile radius

  • of the facility based on the radionuclides released in liquid and gaseous effluents during the time periods of 1 July 1980 through 31 December 1980 and from 1 January 1981 through 31 December 1981. Utilizing the effluent data contained in the

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  • I Semi-Annual Radioactive Effluent Release Reports as source

. 9

. terms, dose equivalent rates were determined using the environmental pathway models specifi,ed in Regulatory Guides 1.109 and 1.111 as incorporated in the NRC computer codes i LADTAP for liquid pathways, and XOQD0Q and GASPAR for gaseous effluent pathways. Dose equivalent rates were calculated for the total body as well as seven organs and/or tissues for adults, teenagers, children, and infants. From the standpoint of liquid effluents, the pathways evaluated included fish ingestion, drinking water, and external exposure to water and sediment.

Based on the findings of the above referenced report, the

, mariaua total body dose from all liquid pathways is received by an adult. Similarly, the maximum total dose to any organ is received by the teen liver. In both of these cases (i.e. ,

adult whole body dose and teen liver dose), 99% and 98%

respectively of these doses were received via the fish 3 ir.gestion pathway.

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In order to determine the dose contribution from the release of liquid effluents, the annual dose to an adult whole body and a

teen liver will be calculated for each of the significant nuclides (see Table 1-1) identified in the liquid weste based l

on the fish ingestion pathway utilizing the following formula:

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,- . . - -.. . - - . . - - . . -- ----- - . - :. _ . - ..... -z -.

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~R apji = 1100U ap MP IB 1p aipj exp (-Ai tp)

D

't F

Where R,pp = total annual dose (per Curie released) to organ "j" of individuals of age group "a" from all of the nuclides "i" in pathway "p" (units = aram/ year - Ci)

U,p = usage factor specifying the exposure time or intake rate for an individual of an age group "a" associated with pathway "p" (units = kg/ year)

Mp =

mixing rates (reciprocal of dilution factor) at the point of exposure or point of haivast (units = dimensionless)

F= flow rate of the liquid effluent (units =

f t /sec) l B =

gp equilibrium bioaccumulation factor for nuclide "i" in pathway "p" (units =

, liters /kg) j  :

I

-w-, ,_...__.__r, , ..- -y.-....,,.,,.m- v.-

D,

=

the dose factor, specific to a specific age group "a", radionuclide "i", pathway "p",

and organ "j", which can be utilized to calculate the radiation dose from an intake of a radionuclide (units = aram/pC1) 1i= radioactive decay constant of nuclide "i" (units = hours -1) t = the average transit time required for nuclides to reach the point of exposure.

For internal dose, t is the total time elapsed between the release of the nuclides and the ingestion of the food and/or water (units = hours) 1100 = factor to convert from (Ci/ year)/(f t /see) to pct / liter-Values for R,pg are contained in Table 1-1. All of these values of R,p3g are on a per Curie basis for each of the nuclides released. Table 1-2 indicates those parameters used l

for selected factors of the formula.

Prior to each radioactive liquid discharge, each liquid waste i

tank to be discharged will be analyzed for isotcpe content utilizing a GeLi detector. On the basis of this analysis, l projected doses to an adult whole body and a teen liver will be calculated using the following relationships:

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R' PSB = 1.05 g apji(WB) i PD = 1.05 g R ,pjggL) C i t

Where PD WB =

Projected dose expected to the whole body 'of an adult due to the release of .the identified concentration of nuclide "i".

(Units = area / year)

Pg = Projected dose expected to the liver of a teenager due to the release of the identified concentration of nuclide "i" (units = sfrea/ year) apji(WB) = t tal annual dose (per Curia released) to the whole body of an adult caused by the ingestion of nuclide "i" (units =

aren/ year-Curia)

R,pjggg) = total annual dose (per Curie rdessed) to the liver of a teenager caused by the ingestion of nuclide "i" (units =

area / year-Curia)

C =

quantity of nuclide "1" identified as f

present in the release (units = Curies)

- 1.05 = correction factor to account for 100% of dose, assuming that 95% of dose received is delivered via the fish ingestion pathway.

The value of 1.05 is used in the equation as a conservative factor to increase the projected dose from an anticipated release to 100%. As long as the 1.05 factor is used, doses received via the drinking water pathway (eg, tritium) need not be accounted for separately.

All projected doses calculated in this manner for each batch of liquid effluent will be summed for comparison with quarterly and annual limits, added to the doses accumulated from other

) .

releases in the quarter and year of interest. In all cases, the following relationships will hold:

For a calendar quarter:

De 11.5 arem total body Dt I "#** # ""I #8"" -

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--t -- r.--~_ -, ,- , ,r ,..,,-,,m.. ,--e- ,.%.~-._ -.- ---__

Far tha calcnder year

- l D < 3.0 mrem total body '

Dg.< 10 mrem for any organ where Dy = total dose received due'to liquid effluent release s '

If these limits are exceeded, a special report will be submitted to the NRC identifying the cause an~d proposed corrective actions. In addition, if these limits are exceeded by a factor of two, calculations shall be made to determine if the dose limits contained in 40 CFR 190 have been exceeded.

Dose limits, as contained in 40 CFR 190 are total body and organ doses of 25 mrem per year and a thyroid dose of 75 arem per year. These calculations will include doses as a result of liquid and gaseous pathways as well as doses from direct radiation. Liquid, gaseous and direct radiation pathway doses will consider the James A. FitzPatrick facility as well as Nine Mile Point Nuclear Station.

In the event the calculations demonstrate that the 40 CFR 190 dose limits, as defined above, have been exceeded, then a report shall be prepared and submitted to the Commission within 30 days as specified in Section 3.6.15.d of the Technical Specifications .

Section 4.0 of the ODCM contains more information concerning calculations for an evaluation of whether 40 CFR 190 limits have been exceeded.

. 3.0 GASEQUS EFFLUENTS 3.1 Setpoint Determinations

. 3.1.1 Basis -

1 Stack gas and off gas monitor setpoints will be established '

such that the instantaneous release rate of radioactive

, materials in gaseous affluents does not exceed the 10 CFR 20 '

4 limits for annual release rate. The setpoints will be activated if the instantaneous dose rate at or beyond the (land) site boundary would exceed 500 area /yr to the whole body or 3000 area /yr to the skin from the continuous release of radioactive noble gas in the gaseous affluent.

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Energency condenser vent monitor setpoints will be established such that the release rate for radioactive materials in gaseous effluents do not exceed the 10 CFR 20 limits for annual release y rate over the projected longest period of release.

Monitor setpointIs from continuous release points will be determined once per quarter under normal release rate conditions and will be based on the isotopic composition of the release and/or a more conservative default composition i

specified in the pertinent procedure. If the calculated 4

setpoint is higher than the existing setpoint, it is not '

+

mandatory that the setpoint be chang:.d.

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m e .r-r-, ,-w.-w y .m,.--.y-.,-w, . -

-...,,.-~.,,,--.,-..c,%~, ,m-~.,-.%- e. e-.,~ , , . . - - - - - .**-,-w. * , - - . - ..- -e-, , - <, - - - - - -

3.1.1 (C nt'd)

Monitor setpoints for emergency condenser vent monitors are conservatively fixed at 5 ar/hr for reasons described in Section 3-1.4 and therefore do not require periodic recalculation.

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2 Under abnormal site release rate conditions, monitor alarm setpoints from continuous release points will be recalculated and, if necessary, reset at more frequent intervals as deemed

. necessary by C&RP Supervision. In particular, contributions from both JAF and the Emergency Condenser Vents shall be assessed.

During outages and until steady state power operation is again realized, the last operating setpoint shall be used.

Since monitors respond to noble gases only, monitor alarm 2

points are set to alarm prior to exceeding the corresponding total body dose rates.

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The skin dose rate limit is not used in setpoint calculations because it is never limiting.

3.1.2 Stack Monitor Satpoints The derr.iled methods for establishing setpoints shall be contained in the station procedures. These methods shall apply l the following general criteria:

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.= . - - - - _ - . - -

3.1.2 Stack Monitor Setpoints (Cont'd)

(1) Rationale for Stack monitor settings is based on the general equation:

release rate, actual = release rate, max. allowable corresponding dose rate, actual corresponding dose rate, max.

allowable EQ g =

(0) _ , ,

X/Q I(QgMg) 500 ar/yr where:

Qg

=

uCi/sec released rate for each isotope, i

, X/Q =

highest land sector site boundary dispersion parameter equal to 1.5 E-6 Sec/m M3 =

gamma air dose factor in units of ar/yr/pci/m3 (see Table 4-1) l (Q) ,,x= instantaneous release rate limit.

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3.1.2 (Cont'd) .

(2) To ensure that 10 CFR 20 and Technical Specification dose rate limits are not exceeded, the bi hi alarms on the stack monitors shall be set lower than or equal to (0.9)

(Q) ,,,. Hi clarms shall be set lower than or equal to (0.5) (Q) .

(3) Based on the above conservatism, the dose contribution from-JAF can usually be ignored. During Emergency Classifi-cations at JAF due to airborne effluent, or after emergency 2 condenser vent releases of significant proportions, the 500 mr/yr value may be reduced accordingly.

d (4) Io convert monitor gross count rates to pCi/sec release rates, the following general formula shall be applied:

4 (C,-B) K, =

Q

" pCi/sec release rate where:

C, = monitor gross count rate in cps or cps B = monitor background count rate l

+

K, = stack monitor efficiency facto:. with units of pCi/sec-cps or pCi/sec-cpm.

i w- - m- .,

3 v. ~ -.,.-y y , ,.-wm._y... - _ - - . _ - - - - . - - - - - - - , . - - _ -

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r 3,1.2 (C:.nt'd)

(5) Monitor K, f actors shall be determined using the general formula:

=

K, Q /(C,-B) where:

Qg =

individual radionuclide stack effluent release rate as determined by isotopic analysis.

K, factors more conservative than those calculated by the above methodology may be assumed.

3.1.3 Recombiner Discharge (Off Gas) Monitor (1) The hi hi alarm points shall activate with recombiner discharge rates equal to or less than 500,000 yC1/sec.

This alarm point may be set equal to or less than 1 Ci/sec for a period of time not to a: ceed 60 days provided the offgas treatment system is in operation.

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3.1.3 (Cont'd)

-(2) The bi alarm points shall activate with recombiner. discharge rates equal to or less than 500,000 pCi/sec (3) To convert monitor mR/hr readings to pCi/sec the formula below shall be applied: .

RK g =

QR pCi/sec recombiner discharge release rate where:

R =

mR/hr monitor indicator s

K R

=

efficiency factor in units of pCi/sec/mR/br determined prior to setting monitor alara points (4) Monitor KR factors shall be determined using the general formula:

( bR" 91/R ,

,where:

l Qg

=

individual radionuclide recombiner discharge release rate as determined by isotopic analysis and flow rate monitor.

I1Rf act rs more conservative than those calculated by the above methodology may be assumed.

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i 3.1.4 Emergency condenser Vent Monitor Satpoint ,

The monitor setpoint was established by calculation (" Emergency I I

Condenser Vent Monitor Alarm Setpoint", January 13,1986, NMPC File Code #16199). Assuming a hypothetical case with (1) reactor water iodine concentrations higher than the Technical 4

Specification Limit. (2) reactor water noble gas concentrations e

higher than would be expected at Technical Specification iodine 2

levels, and (3) leakage of reactor steam into the emergency 6

condenser shell at 300% of rated flow (or 1.3 x 10 lbs/hr),

the calculation predicts an emergency condenser vent monitor response of 20 mr/hr. Such a release would result in less than 1

10 CFR 20 dose rate values at the site boundary and beyond for typical emergency condenser cooldown periods.

'^ t Since a 20 ar/hr monitor response can, in theory, be achievable 1

only when reactor water iodines are higher than permitted by .

Technical Specifications, a conservative monitor setpoint of 5 ar/hr has been adopted.

t-3.2 Dose and Dose Rate Determinations i

In accordance with specification 4.6.15.b.(2) and 4.6.15.b.(3), dose and dose rate determinations will be made monthly in order to p determine:

(1) Total body dose rates and gamma air doses at the maximum X/Q

, 4 1and sector site boundary interface and beyond.

1

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4

. _ . . _. _. ~ . . _ _ _ . . _ _ - - .. - _-

l 3.2 (Cont'd)

(2)

Skin dose rates and beta air doses at the maximum X/Q 1and .

sector site boundcry interface and beyond.

(3) The critical organ dose and dose rate at the maximum X/Q 1and sector site boundary interface and at a critical receptor i

location beyond the site boundary.

Either average meteorological data (ie, marimum five year annual

. average X/Q and D/Q values in the case of elevated releases or 1985 annual average X/Q and D/Q values, in the case of ground level 2l '

releases) or real time meteorological data shall be utilized for dose and dose rate calculations. Where average meteorological data is '

assumed, dose and dose rates due to noble gases at locations beyond the site boundary will be lower than equivalent site boundary dose i

and dose rates. Therefore, toder these conditions, calculation of noble gas dose and dose rates beyond the =mwimum X/Q 1and sector site boundary locations can be neglected.

The frequency of dose rate calculations will be upgrcded when elevated releas'e rate conditions specified in subsequent sections 3.2.1.1 and 3.2.1.2 are realized.

Emergency condenser vent release contributions to the monthly dose and dose rate determinations will be considered only when the 2

emergency condenser return isolation valves have been opened for reactcr cooldown or if Emergency Condenser tube leaks develop with or without the system's return isolation valve opened. Without tube leakage or opening of the return isolation valves, releases from this system are negligible and the corresponding dose contributions do not i

have to be included.

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-__.-_A . - , - - - - . - - , , ,

I 3.2 (Cont'd)

When releases from the emergency condenser have occurred, dose and

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2 dose rate determinations shall be performed using methodology in 3.2.1 and 3.2.2. Furthermore, environmental sampling may also be initiated to refine any actual contribution to doses. See Section 3.4.

Critical organ doses and dose rates may be conservativaly calculated '

by assuming the aristence of a so-called " moving" critica1 receptor.

At this " moving" critical receptor location, it is assumed that all pathways are applicable and the bigbest X/Q and/or D/Q value for actual pathways as noted in Table 3-1 are in effect. A person's dose at the " moving" critical receptor location is equal to the same dose -

that person would receive if they were simultaneously subjected to the highest pathway dose at each critical receptor identified for each pathway.

If dose or dose

  • rates calculated, using the assumptions noted above, reach Technical Specification limits, actual pathways will be evaluated, and doses / dose rates shall be calculated at separate critical receptor locations and compared with applicable limits.

Not all pathways need be considered in dose and dose rate calculations at each critica1 receptor location. For example, when calculating land sector site boundary doses and dose rates for particulates, iodines and tritium, only the ground deposition and inhalation pathways apply.

3.2.1 Dose Rate 3.2.1.1 Noble Gasas In accordance with the provisions of 10 CFR 20 the dose rates from noble gas release from the site to unrestricted areas arc to be limited to 500 mres/yr to the total body and 3,000 ares /yr to the skin. Dose rate calculation will be' performed monthly, or when the Hi Hi stack monitor alarm point is reached, using the following equations:

For total body dose rates (in ar/yr):

DR = 3.17 x 10" Mg (I/Q)Qg /sec For skin dose rates:

-8 DR ,3 =

[3.17 x 10 N1(x/Q)Qg/sec) + DR where:

Mg , N g, I/ Q , Q g, 3.17 x 10 are as defined in

. section 3.2.2.1

- - -=* - = =

. , - - ._ s. -r------

. . . . - - . . . . - .~. .- ..--. .- .

3.2.1.2 Tritium, Iodines and Particulates (1) The dose rate in unrestricted areas from the release of tritium, iodine-131, iodine-133 and all radionuclides in particulate form with half lives grerter than 8 days is limited to 1500 mren/ year to any organ.

(2) In order to ensure that the 1500 mren/ year dose rate limit is not exceeded, particulate, iodine and tritium off site dose rate calculations shall be performed monthly and whenever particulate and iodine relesse

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rates exceed 10 pCi/see using the equation given in Section 3.2.2.2 with Q expressed in pCi/sec.

-2 When the release rate exceeds 10 uC1/see, the dose rate assessment shall also include JAF contribution.

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(3) ~ The use of the 10 pCi/sec release rate threshold to perform dose rate calculations is justified as follows: The 1500 arem/yr organ dose rate limit corresponds to a minimum release rate limit of 0.27 uCi/sec calculated using the equation:

1500 = (Q/sec) (R,W}),,

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l

... . .. . - . . . - . . - - - - . .  := . . .- . ..- .. . . . .

3.2.1.2 (Cont'd) where:

1500 =

site boundary dose rate limit in aram/ year (Rg g),,, =

the maximum curie to dose conversion factor equal to 5500 arem-sec/pci-yr for Sr-90, child bone at the " moving" critical receptor location beyond the site boundary. ,

3.2.2 Dose Calculations will be performed monthly at a minimum, ts demonstrate that doses resulting from the release of noble gases, tritium, I-131 I-133 and particulates with half lives greater than 8 days are within the limits specified in 10 CFR 50, Appendix I. These limits are:

Noble Gas Air Dose

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5 mr gamma / calendar quarter 10 mrad beta / calendar quarter 10 mr gamma / calendar year 20 mrad beta / calendar year Radioiodines Tritium & Particulates 7.5 mrem to any organ / calendar quarter 15 mrem to any organ / calendar year H r . - .---_ _ _ _ _ _ _ __-__ _ _ _ _ _ _

. =-.. . . .

=: .-- -. := -

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  • 3.2.2.1 Noble Gas Air Dosa The air dose at the critical receptor due to noble gas r releases is determined as follows:

For gamma radiation D = 3.17 x 10 -8 zg i gjg g i For beta radiation D,= 3.11 x 10 IN gX/Q Qg where Mg = air gamma dose factor in (ar/ year per uCi/m3 ) for each isotope 1 (Table 4-1)

Ng =

air beta dose factor (arad/ year per 3

uC1/m ) for each isotope 1 (Table 4-1)

X/Q = the relative plume concentration (in units of sec/m ) at the land sector site boundary or beyond. Either average meteorological. data (Table 3-1 or Appendix C), or real time values may be assumed. 2. '

" Elevated" X/Q values are used for stack  !

releases; " Ground" X/Q values are used for

, Emergency Condenser Vent releases.

Qg '= the total quantity of isotope i released during the period, (uci)  !

3.17 X 10-8 = the inverse of the number of seconds in a year 4 3 -- y.-e,. s- - - - - -

.' . 1

, 3.2.2.2 Radioiodine, Tritium & Particulates  !

The doses to an %dividual from I-131, I-133, tritium, and particulates s vb balf lives greater than 8 days i

will be calculated as follows:

Dose = 3.17 I 10-8 I I R S g ijak ji Where W3 = dispersion parameter either I/Q (sec/m ) or 2

D/Q (1/m ) depending on pathway and receptor

,, location assumed. Either average meteorological 2 data (Table 3-1, or Appendix C) or real . time values may be assumed. " Elevated" W values 3

are used for stack releases; " Ground" V values 3

are used for Emergency Condenser Vent releases.

Qg = the total quantity of isotope i released during the period, (uci)

  • R ijak

=

the dose factor for each isotope i, pathway j, age group a, and organ k (Table 4-2, through 4-20)

-8

3. 17 'I 10 = the inverse of the number of seconds in a year The R values contained in Tables 4-2 through 4-20 were calculated using the methodology defined in NUREG-0133 l

1 and Regulatory Guide 1.109 Revision 1.

l 1

l l

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l _ _ ._ _ _ _ _ _ .. .

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3.2.2.3 Accumuisting Doses Doses will be cElculated monthly, at a minimum, for gamma air and beta air, the identified critical organ, 4

and age group. Results will be summed for each calendar quarter and year.

It has been historically demonstrated that the critical pathway is usually the grasnow milk pathway i

and the critical organ is the infant's thyroid.

f For this reason, monthly infant thyroid dose estimates will normally be made prior to receipt of all analysis data (i.e. , strontium and tritium). The critical doses are based on the following pathways.

noble gas plume air dose ground plane dose (deposition) '

inhalation dose cow's milk dose

- goat's milk dose meat consumption dose vegetation (food crops) dose The quarterly and annual results shall be compared to the limits listed in paragraph 3.2.2. If the limits are exceeded, special reports, as required by Section 6.9.3 of the Technical Specification, shall be submitted.

-2 6 -

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n-eer =.-vr- 4 --

  • _ ---n -

-.,-,.-,p- - , . _

3.3 Critical Receptors In accordance with the provisions of 10 CFR 20 and 10 CPR 50, Appendix I, the critical r'eceptors have been identified and are '

contained in Table 3-1.

For noble gas doses, one of two critical receptor locations will be assumed. When maximum five year average annual X/Q values are used, the critical receptor is the maximum I/Q 1and sector site boundary interface. When real time meteorological X/Q values are used, the critical receptor may either be the maximum X/Q 1and sector site boundary location, or the downwind location of greatest X/Q residence (e.g.,1.5 miles east), whichever is higher.

For I-133, I-131, tritium and particulate radionuclides with a half life of greater than eight days, the critical pathways are milk (cow and goat), meat, vegetation, inhalation and direct radiation (ground plane) as a result of ground deposition.

De cow milk and goat milk pathway will be based on the greatest D/Q milk cow and milk goat location as determined by technical b

specification 3.6.22. De inhalation dose pathway will be based on the greatest X/Q residence as determined by technical specification i 3.6.22 since this location would have the greatest potential occupancy time. The ground plane dose pathway will be calculated as the greatest D/Q residence because of the greatest potential occupancy time.

3[3 For the meat consumption dose pathway, the critical receptor is the greatest D/Q mest animal location. This location has been determined in conjunction with the land use census (technical specification

- . .. - ~

3.3 (Cont'd) 3.6.22) and is subject to change. The vegetation (food crop) dose is based on the greatest D/Q garden location from which samples are

^

taken. This location also may be modified as a result of vegetation sampling surveys.  ;

4 3.4 - Refinement of Offsite Doses Resulting from Emergency Condenser Vent Releases The doses resulting from the operation of the emergency condensers and calculated in accordance with 3.2.2 may be refined using data from actual environmental samples.

Ground deposition samples will be obtained from an area or areas of maximum projected deposition. These areas are anticipated to be at or 2 near the site boundary and near projected plume centerline. Using the methodology found in Regulatory Guide 1.109, the dose will be calculated to the maximum exposed individual. This dose will then be compared to the dose calculated in accordance with 3.2.2. The comparison will result in an adjustment factor of less than or greater l

l than one which will be used to adjust the other doses from other pathways.

I i

Other environmental samples may also be collected and the resultant celculat.sd doses to the maximum exposed individual compared to the l dose calculated per 3.2.2. Other environmental sample media may l

include milk, vegetation (such as garden broadleaf vegetables), etc.

t

! The adjustment factors from these pathways may be applied to the doses

~

calculated per 3.2.;! on a pathway by pathway basis or several pathway adjustment factors may be averaged and used to adjust calculated doses.

l l _ _ . _ _ _ --

l- .

. 3.4 (Cont'd) , i Doses calculated from actual environmental sample media will be based on the methodology presented in Regulatory Guide 1.109. Th e 2 ,

regulatory guide equations may be slightly modified to account for short intervals of time (less than one year) or modified for simplicity purposes by deleting decay factors. Deletion of decay factors would yield more conservation results.

4.0 40 CFR 190 REQUIRDENTS The " Uranium Fuel Cycle" is defined in 40 CFR Part 190.02 (b) as follows:

i

" Uranium fuel cycle means the operations of milling of uranium ore, chemical conversion of uranium, isotopic enrichment of uranium, fabrication of uranium fuel, generation of electricity by a light-water-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 for public use utilizing .

nuclear energy, but excludes mining operations, operations at waste disposal sites, transportation of any radioactive material in support of these operations, and the reuse of recovered nonsranium special nuclear and by-product materials from the cycle."

Section 3.6.15.d of the Technical Specifications requires than when the calculated doses associated with the affluent releases exceed twice the limits of sections 3.6.15.a.(2)(b), 3.6.15.b.(2)(b) and 3.6.15.b.(3)(b),

then calculations shall be made including direct radiation contributions from the reactor units and outside storage tanks (as applicable) to l determine whether the 40 CFR 190 dose . limits have been exceeded.

P

~

n.,. - . . . . . - . ~ , --m--o- o~ - . ~ ~ ~ - - * * - - ~

4.0 (Cont'/.) \

If such is the case, Niagara Mohawk shall submit a Special Report to the NRC and limit subsequent releases such that the dose commitment to a real individual from all uranium fuel cycle sources is limited to 125 area to the total body or any organ (except the thyroid, which is limited to 175 arem) over the calendar year. This report is to demonstrate that radiation exposures to all real individuals from all uranium fuel cycle sources (including all liquid and gaseous effluent pathways and direct ,

radiation) are less than the limits in 40 CFR Part 190. If releases that result in doses exceeding the 40 CFR 190 limits have occurred, then a variance from the NRC to pernit such releases will be requested'and if possible, action will be taken to reduce subsequent releases.

The report to the NRC shall contain:

1) Identification of all uranium fuel cycle facilities or operations within 5 miles of the nuclear power reactor units at the site, that contribute to the amual 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 existing pathways and sources of radioactive affluents and direct radiation.

The maximum total body and organ doses resulting from radioactive matertal

+.

in liquid effluents from Nine Mile Point Unit 1 will be summed with the maximum doses resulting from the releases of noble gases, radioiodines, i

and particulates for the other calendar quarters (as applicable) and from the calendar quarter in which twice the limit was exceeded. The direct dose components will be determined by either calculation or actual measurement.

L. _ _ _ _ _ _ . _ _ _ __-. _ . - . . __

4.0 (Cont'd) .

The doses from Nine Mile Point Unit 1 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.

For the purpose of calculating doses, the results of the Radiological Environmental Monitoring Program may be included for providing 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, the fish consumption and shoreline sediment ground dose will be considered. Since the doses from other aquatic pathways are 2 insignificant, fish consumption and shoreline sediment are the only two pathways that will be considered. The dose associated with fish consumption may be calculated using the ODCM methodology or by calculating a dose to man based on actual fish sample analysis data.

I The dose associated with shorelire sediment is based on the l

i assumption that the shoreline would be utilized as a recreational area. This dose may be derived from liquid effluent data or from actual shoreline sediment sample analysis data.

Doses. to members of the public from the fish consumption and l

shoreline sediment pathways will be calculated using Regulatory Guida l 1.109 methodology or ODCM methodology.-

I l

l l l

4.2 Evaluation of Doses Prom Gaseous Effluents For the evaluation of doses to real members of the public from gaseous affluents, the pathways contained in section 3.2.2.3 of the  ;

ODCM will be considered. 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 3.2.2.3 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.

4.3 Evaluation of Doses From Direct Radiation Section 3.6.15.d of the Technical Specifications requires that the dose contribution as a result of direct radiation be considered when evaluating whether the dose limitations of 40 CFR 190 have been exceeded. Direct radiation doses as a result of the reactor, turbine and radwaste buildings and outside radioactive storage tanks (as applicable) may be evaluated by engineering calculations or by evcluating environmental TLD results at critical receptor locations, site boundary or other special interest locations.

. 4.4 Doses to Hembers of the Public Within the Site Boundary.

Section 6.9.1.e of the Nine hile Point Unit 1 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 5.1-1 of the specifications. A member of the public, as defined by the Technical Specifications, would be represented by an individual who visits the sites' Energy Information Center for the purpose of observing the educational displays or for picnicing and associated activities. It is conservatively assumed that an individual would spend four hours 2 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 i

summer season is the period of concern. The second report will address this dose based on forty eight hours occupancy. ~ 0ther time 2

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 then 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

. 4.4 (Cont'd)'

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, include several pathways. These include: the direct radiation gamma dose to an individual from an overhead plume, .a submersion gamma plume dose, and a ground plane dose (deposition). Other pathways, such as the ingestion pathway, are not applicable. In addition, pathways associated with water i

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 I/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.

gNgF X/Q m g3, R, tl R =

e

. 4.4 (Cont'd) where )

R =

the maximum dose for the period in question to the lung (j) for all radionuclides (1.) for the adult age group (a) in mrem per time period.

C=g The average concentration in the stack release of radionuclide i in pCi/m for the period in question F = Unit 1 average stack flowrate in a /sec.

X/Q = The plume dispersion parameter for a location 0.50 niles west of NMP-1 (The plume dispersion parameter is 1.176E-07 and was obtained from the C.T. Main five year average grazing season X/Q tables. A X/Q value based on real time meteorology may also be utilized for the period in question, if desired) .

DF ,

=

the inhalation dose factor for radionuclide i, the lung j, and adult age group a in mrem per pCi found on Table E-8 of Regulatory Guide 1.109. 2 R,

=

annual air intake for individuals in age group a in M per year (this value is 8,000 m 3per year and was obtained from Table E-5 of Regulatory Getde 1.109).

t =

fractional portion of the year for which radionuclide i was detected and for which a dose is to be 2

calculated (equals 0.0055 years) .

. 4.4 (Cont'd)

The direct radiation gamma dose pathway includes any gamma doaes from an overhead plume, submersion in the plume and' ground plane dose (deposition). This general pathway will be evaluated by average environmental TLD readings. At least two environmental TLD locations will be utilized and located in the approximate area of the Energy Information Center (EIC) and the facility picnic area. These TLDs will be placed in the field on approximately July 1 and removed on approximately September 30 of each year (this time interval is composed of one quarterly TLD collection period). The average TLD readings will be adjusted by the average control TLD readings. This is accomplished by subtracting the average quarterly control TLD

-tue from the average EIC TLD value. The applicable quarterly control TLD va - will be utilized after adjusting for the appropriate time period (ai. oplicable)..

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. 2 The Radiological Environmental Monitoring Program is a joint effort between the Niagara Mohawk Power Corporation and the New York Power Authority, the owners and operators of the Nine Mile Point Unit 1 and the James A. FitzPatrick Nuclear Power Plant, respectively. Sampling locations are chosen on the basis of historical average dispersion or deposition parameters from both units.

s.

y -.-.,3 y .--9.- . w .9--%

. 5.1 (Cont'd)

The average dispersion and deposition parameters for the two units have been calculated for a 5 year period,1978 through 1982. These dispersion calculations are attached as Appendix C. The calculated dispersion or deposition parameters will be compared to the results of the annual land use census. If it is determined that a nilk sampling location exists at a location that yields a significantly higher (e.g. 50%) 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 Commirision approved or sponsored Interlsboratory Comparison Program, such as the EPA Crosscheck Program. Participation shall be only for those media, ,

t e.g. , sir, milk, water, etc. , that are included in the Nine Mile i

Point Environmental Monitoring Program and for which cross check samples are available. The site identification symbol or the actual Quality Control sample results shall be reported in the Annual Radiological Environmental Operating Report so that the Commission i

j staff may evaluate the results.

l l

l l l

_ _..- .. , - . . . _ - 5 .? _' _ - - -

- -.-. :==

, - 5.2 (Cont'd)

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 5.3 Capabilities for Thermoluminescent Dosimeters Used for Environmental Measurements Required detection capabilities for thermoluminescent dosimeters used for environmental measurements required by Table '4.6.20-1, footnote b

$- of the Technical Specifications are based on ANSI Standard N545, section 4.3. TLDs are defined as phosphors packaged for field use.

In regard to the detection capabilities for thermoluminescent l

l dosimetersi 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:

l l

- ^ - -

4

~

  • 5.3.1 Uniformity shall be determined by giving TLDs from the same i

batch an exposure equal to ths.t resulting from an exposure '

rate of 10 uR/hr during the field cycle. The responses obtained shall have a relative standard deviation of less than 7.5%. A total of at least 5 TLDs shall be evaluated.

5.3.2

~

Reproducibility shall be determined by giving TLDs repeated exposures equal to that resulting from an exposure rate of 10 uR/hr during the field cycle. The average of the relative standard deviations of the responses shall be less than 3.0%. A total of at least 4 TLDs shall be evaluated.

, 5.3.3 Dependence of exposure interpretation on the length of a

fielu cycle shall be examined by placing TLDs for a period equal to at least a field cycle and a period equal to half the same field cycle in an area where the exposure rate is known to be constant. This test shall be conducted under approximate average winter temperatures and approximate average summer temperatures. For 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 t

l. 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 i

obtained with the calibration source by more than 25% for I photons with energies greater than 80 kev and shall not be i

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.

_ _ , _ _~ . __.__. ~.. __

- , . . - - ~ . . . . _ .-- - - - =:  :: -

l l

)

5.3.5 The directional dependence of the TLD response. shall be l

determined by comparing the response of the TLD exposed in-the routine orientation with respect to the calibration source with the response obtained for different orientations. To accomplish this, the TLD shall be rotated

'through at least two perpendicular planes. The response averaged over all directions shall not differ from the response obtained in the standard calibration position by more than 10%. A total of at least 4 TLDs shall be evaluated.

5.3.6 Light dependence shall be determined by placing TLDs in the field for a period equal to the field cycle under the four conditions found in ANSI N545, section 4.3.6. The results obtained for the unwrapped TLDs shall not differ from those obtained for the TLDs wrapped in aluminum foil by more than 10%. A total of at least 4 TLDs shall be evaluated for each of the four conditions.

5.3.7 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 conditicus: (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, l as appropriate, shall be dried before readout. The

5.3.7 (Cont'd) response of the TLD exposed in the plastic bag containing water shall not differ from that exposed in the regular plastic bag by ac' a than 10%. A total of at least 4 TLDs shall be evaluated for each condition.

i 5.3.8 Self irradiation shall be determined by placing TLDs for a period equal to the field cycle in an area where the exposure rate is 1ers 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.

I 4

TABLE 1-1 Rapji VALUES FOR THE NINE MILE POINT UNIT 1 FACILITY ADULT - TOTAL BODY TEEN - LIVEL NUCLIDE kapji (MREM /YR-C1) Rg (MREM /YR-Ci)

.89S r 2.05 E - 3 N/A 90Sr 4.37 E - 1 N/A 13Q, 1.89 E + 0 ~2.35 E + 0 137Cs 1.12 E + 0 1.78 E + 0 58Co 6.4 7 E - 4 2.87 E - 4 60Co 1.85 E - 3 ' 8.38 E - 4 54h 2.72 E - 3 1.19 E - 2 9

e d

~ ,

TABLE 1-2 PARAMETERS FOR THE LIQUID EFFLUENT PATHWAY REFERENCE PARAMETER VALUE (REG. GUIDE 1.10 9)

.Unp Adult = 21.0 Kg/yr Table E-5 Teen = 16.0 Kg/yr Mp 0.2 Site Specific F 590 ft.3/second Site Specific Bip Each element Table A-1 41pi Each radionuclide Tables E-11 to E-14 t p. ~ 26.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Site Specific l .

s e

TABLE '2-1 Average Energy Per Disintegration ISDTOPE E mev/ dis (Ref) EBaev/ dis (4) (Ref)

Ar-41 1.294 (3) 0.464 (3)

Ke-83m 0.00248 (1) 0.0371 (1)

Kr-85 0.0022 (1) 0.250 (1)

Kr-85m 0.159 (1) 0.253 (1) 4 Kr-87 0.793 (1) 1.32 (1)

Kr-88 1.95 (1) 0.377 (1)

'Kr-89 2.22 (2) 1.37- (2)

Kr-90 2.10 (2) 1.01 (2)

Xe-131M 0.0201 (1) 0.143 (1)

Xe-133 0.0454 (1) 0.135 (1)

Xe-133n 0.042 (1) 0. 19 (1)

Xe-135 0.247 (1) 0.317 (1)

Xe-135m 0.432 (1). 0.055 (1)

Xe-137 0.194 (1) 1.64 (1)

Xe-138 1.18 (1) 0.611 (1)

(1) ORNL-4923, Radioacrive Atoms - Supplement I, M.S. Martin, November 1973.

(2) NEDO-12037, " Summary of Gamma and Beta Emitters and Intensity Data"; M.E.

Meek, R.S. Gilbert, January 1970. (The average energy was computed from the maximum energy using the ICRP II equation, not the 1/3 value assumption used in this reference).

(3) NCRP Report No. 58, "A Handbook of Radioactivity Measurements

! Procedures"; 1978 (4) The average energy includes conversion electrons.

l l l

u i .

4

  • TABLE 3-1 Critical Receptor Dispersion Parameterel For Ground Level and Elevated Releases ELEVATED ELEVATED GROUNDS GROUND 5 1 see LOCATION DIR MILES I/O a+se D/0 af Il m3 g D ]a Rssidences E (98*) 1.4 1.8 E-07 2 5.2 E-0 9 2

3 Dairy Cows SE (130') 2.6' 2.2 E-08 7.0 E-10 3 Milk Goats 1.3 E-08 3 1.6 E-10 3 E (88') 7.9 Mast Animal s ESE (115 *) 1.8 5.1 E-08 3 1.7 E-0 9 3 ,

Gardcns 3 3 E (97') 1.8 -

1.0 E-07 3.5 E-0 9 Sita Boundary ENE (67 *) 0.4 2.4 E-062 ,4 4.4 E-082 ,4 ,

1. These values will be used in dose calculations beginning in April 1986 but may be revised periodically to account for changes in i locations of farms, gardens or critical residences.
2. Values based on 5 year annual meteorological data (C.T. Main, Rev. 2)
3. Values based on 5 year average grazing season meteorological data (C.T. Main, Rev. 1) 4.- _Value are based on most restrictive X/Q 1and-based sector (ENE).

(C.T. Main, Rev. 2)

5. Values (to be available by April 1,1986) are based on average annual
. meteorological data for the year 1985. In the interim, ground level r

X/Q and D/Q values will conservatively be set to equal 10X the corresponding Elevated values.

I e

45-e.

e ,-9 .

% -y. , ._ ,, - - , - . ._ m -y- w

. . . . = . .- - - -- . . - - - . -

.~.-

, TABLE 4-1 DOSE FACTORS FOR NOBLE GASES Gamma Air Beta Air l Dose Factor Dose Factor Ni Ni a r-et. 3

~ aradM ~

i Radionuclides pCi yr pCi-yr Kr-83a 1.93E+01 2.88E+02 Kr-85a 1.23E+03 1.97E+03 Kr-85 1.72E+01 1.95E+03 Kr-87 6.17E+03 1.03E+04 Kr-88 1.52E+04 2.93E+03 Kr-8_9 1.73E+04 1.06E+04 Kr-90 1.63E+04 7.83E+03 Xe-131m 1.56E+02 1.11E+03 Xe-133a 3.27E+02 1.4 8E+03 Xe-133 3.53E+02 1.05E+03 Xe-135m 3.36E+03 7.39E402 Xe-135 1.92E+03 2.46E+03 Xe-137- 1.51E+03 1.27E+04 Xe-138 9.21E+03 4.75E+03 Xe-139 5.28E+03 6.52E+04 Ar-41 9.30E+03 3.28E+03 t

i Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," dated October 1977, page 1.109-21.

i I

s TABLE 4-2 R VALUES - QOW'S MILK - INFANT f a #-arem/y r uC1/sec

. 1 NUCLIDE T. BODY GI-1RACI B0hE LIVER KID 8eEi 1hikO1L Lbh6

, H 3* 2.40E 03 2.40E 03 2.40E 03 2 40E 03 2.40E 03 2.40E 03 Cr 51 7.46E 04 2.17E 06 1.06E 04 4.67E 04 4.47E 04 Hn 54 4.54E 06 7.36E 06 2.00E 07 4.44E 06 Fe 59 7.21E 07 8.74E 07 1.05E 08 1.83E 08 5.41E 07, Co 58 2.88E 07 2.88E 7 1.15E 07 Co 60 1.11E 08 1.12E 08 4.71E 07 Zn 65 5.26E 09 9.63E 09 3.32E 09 1.14E 10 5.53E 09 Sr 80 1.70E 08 1.22E 08 5.94E 09 '

Sr 90 1.79E 10 8.75E 08 7.01E 10  :

Zr 95 5.58E 02 3.92E 05 3.23E 03 7.87E 02 '6.46E 02 1 131 6.92E 08 5.62E 07 1.34E 09 1.57E 09 1.84E 09 5.17E 11 1 133 7.91E 06 4.57E 06 1.85E 07 2.70E 07 3.17E 07 4.91E 09 j Cs 134 3.59E 09 9.65E 07 1.90E 10 3.55E 10 9.14E 09 3.75E 09 Ca 136 1.03E 09 4.19E 07 9.37E 08 2.76E 09 1.10E 09 2.25E 06 Cs 137 2.37E 09 '1.04E 08 2.85E 10 3.34E 10 8.96E 09 3.63E 09 l

Ba 140 5.94E 06 2.83E 07 1.15E 08 1.15E 05 2.74E 04 7.03E 04 l .

n' l Ce 141 1.44E 03 6.30E 06 2.00E 04 1.22E 04 3.76E 03

( .

l

  • mrea/yr uCi/m 3 i

e TABLE 4-3 R VALUES - COW'S MILK - CHILD m 2-ares /yr -

uC1/sec NUCLIDE T. BODY GI-1RACT BONE LIVIA K1bhhi .ThYkulb LbhG H 3* 1.58E 03 1.58E 03 1.58E 03 1.58E 03 1.58E D3 1.58E 03 1.58E 03 1 Cr 51 4.71E 04 2.50E 06 7.14E 03 2.61E 04 4.77E 04 f

Mn 54 2.87E 06 9.04E 06 1.08E 07 3.02E 06 ,

Fe 59 4.52E 07 9.45E 07 5.61E 07 9.08E 07. 2.63E 07, Co 58 1.77E 07, 3.37E 07 5.77E D6 {

i Co 60 6.81E 07 1.28E 08 2.31E 07 Zn 65 4.10E 09 1.16E 09 2.47E 09 6.59E 09 4.15E 09 j i

,j Sr 89 8.93E 07 1.21E 08 3.13E 09 f it l Sr 90 1.63E 10 8.68E 08 6.44E 10 Zr 95 3.56E 02 4.17E 05 1.82E 03 4.00E 02 5.72E 02 1 131 3.66E 08 5.73E 07 6.40E 08 6.44E 08 1.06E 09 2.13h 11

~

1 133 4.11E 06 4.38E 06 8.78E 06 1.09E 07 1.81E 07 2.02E 09 i

ca 134 4.09E 09 1.05E 08 1.18E 10 1.94E 10 6.01E 09 2.16E 09 Ca 136 ,

8.53E 08 4.63E 07 4.80E Ob- 1.32E 09 7.07s 06' 1.b>E 06 Cs 137 2.52E 09 1.07E 08 1.79E 10 1.71E 10 5.57E 09 2.00E 09 Ba 140 3.27E 06 2.84E 07 5.60E 07 4.91E 04 1.60E 04 2.93E 04 l Ce 141 7.47E 02 6.28E 06 1.01E 04 5.03E 03 2.21E 03 l 1

  • ares /yr' l uC1/mJ .,

li

l.

, TABLE 4-4 R VALUES - COW'S MILK - TEEN abrem/yr uC1/sec

^

NUCLIDE T. BODY Cl-1RACI BONE LIVER KIDhE7 1h1 KOIL LUNG

! H 3* 1.00E 03 1.00E 03 1.00h 03 1.00h 03 l'.00h 03 1.00E 03 -

)

Cr 51 2.31E 04 3.88E 06 $.06h 03 1.26E 04 3.30L 04 l t

Hn 54 1.43E 06 1.48E 07 7.20h 06 2.15E 06-Fe 59 2.18E 07 1.34E 08 2.42E 07 5.65h 07 1.76E 07

, Co 58 8.70E 06 5.21E 07 3.78E 06  !

4 Co 60 3.35E 07 1.94E 08 1.49E 07 Zn 65 2.04E 09 1.85E 09 1.26E 09 4.38E 09 2.80E 09 Sr 89 3.62E 07 1.50E 08 1.26E 09 Sr 90 9.42E 09 .1.07E 09 3.81E 10 .

. . Zr 95 1.70E 02 5.70E 05 7.83E 02 2.47E 02 3.63E 02 1 131 1.98E 08 7.31E 07 2.64E 08 3.69E 08 6.36E 08 1.08E 11' I 133 1.87E 06 4.64E 06 3.61E 06 6.13E 06 1.06E 07 8.56E 08 'f I

Cs 134 5.60E 09 1.50E 08 5.12E 09 1.21E 10 3.83h 09 1.46E 09 Cs 136 5.62E 08 6.73E 07 2.13E 08 8.37E 08 4.55E 08 ).lbE 07 l l Cs 137 3.44E 09 1.40E 08 7.42E 09 9.87E 09 3.36E 09 1.30h 09 -

4 i l Ba 140 1.58E 06 3.58E 07 2.32E 07 2.84E 04 9.65h 03 1.91E 04 Ce 141 3.14E 02 7.83E 06 4.10E 03 2.74E 03 1.29h 03 3

  • mtem/yr 3

uC1/mJ >

49

.t TABLE 4-5 R VALUES - COW'S MILK - ADULT a b res/yr uC1/sec j NUCLIDE T. BODY Cl-1RACT BONE LIVER KlbhEY 1hYR01D LUhG H 3* 7.69E 02 7.69E 02 7.69E 02

  • 7.69E 02 7.69E 02 7.,69E 02 Cr 51 1.32E 04 3.32E 06 2.91E 03 7.90E 03 1.75E 04 ,

i Hn 54 - 8.25E 05 1.32E 07 4.32E 06 1.29E 06

(

Fe~59 1.25E 07 1.09E 08 1.39E 07 3.26E 07 9.lbE 06 i

Co 58 5.03E 06 4.56E 07 2.24E 06 l.

Co 60 1.93E 07 1.65E 08 8.77E 06 Zn 65 1.18E 09 1.65E 09 8.21E 08 2.81E 09 1.75E 09 Sr 89 1.97E 07 1.10E 08 6.85E 08  !

Sr 90 6.62E 09 7.80E 08 2.70E 10 Zr 95 9.72E 01 4.55E 05 4.48E 02 1.44E 02 2.25E 02 1 131 1.19E 08 5.49E 07 1.45E 08 2.08E 08 3.57E 08 6.82E 10 1 133 1.05E 06 3.09E 06 1.98E 06 3.44E 06 6.01E 06 5.06E 08 Cs 134 5.74E 09 1.23E 08, -2.95E 09 7.02E 09 2.27E 09 7.54E 06 Ca.136 3.55E 08 5.60E 07 1.25E 08 4.93E 08 2.74E 08 3.76E 07 Cs 137 3.65E 09 1.08E 08 4.09E 09 5.59E 09 1.90E 09 6.31E 06 Ba 140 8.43E 05 2.65E 07 '1.29E 07 1.62E 04 5.49E 03 9.2SE b3 Ce 141 1.71E 02 5.78E 06 '.24E 03 1.51E 03 7.02E 02

  • mres/yr '

uC1/mJ

s.

o

  • TABLE 4-6 R VALUES - COAT'S HILK - INFANT m2 -erem/yr uC1/sec NUCLIDE T. BODY GI-TRACT BONE LIVIJL K1bhEY 1hYR01D LUNG H 3* 4.90E 03 4.90E 03 4.90E 03 4.90E 03 4.90E 03 4.90E 03 Cr 51 8.95E 03 2.61E 05 1.28E 03 5.84E 03 '1.14E 04 Mn 54 5.45E 05 8.83E 05 2.40E 06 5.33E 05 Fe 59 9.37E 05 1.14E 06 1.36E 06 2.3bE 06 7.0SE 05, f

Co 58 '3.45E 06 3.45E 06 1.38E D6 Co 60 1.34E 07 1.35E 07 '

5.65E 06 Zn 65 6.31E 08 1.16E 09 3.99E 08 1.37E 09 6.63E 08 Sr 89 3.58E 08 2.57E 08 1.25E 10 Sr 90 3.57E 10 1.84E 09 1.47E 11 l Zr 95 6.70E 01 4.70E 04 3.88E 02 9.45E 01 1.02E 02 I 131 8.31E 08 6.74E 07 1.60E 09 1.89E 09 2.21E 09 6.21E 11 1 133 9.49E 06 5.48E 06 2.23E 07 3.24E 07 3.81E 07 5.89E 09 Cs 134 1.08E 10 2.89E 08 5.71E 10 1.07E 11 2.74E 10 1.12E 10 Ca 136 3.09E 09 1.26E 08 2.81E 09 8.27E 09 3.30E 09 6.74E 08 Cs 137 7.10E 09 3.13E 08 8.55E 10 1.00E 11 2.69E 10 1.09E 10 Ba 140 7.13E 05 3.40E 06 1.38E 07 1.38E 04 3.29E 03 6.5bE 03 Ce 141 1.72E O2 7.57E 05 2.40E 03 1.46E 03 4.52E 02 l'

  • mrem /yr UC1/mJ 3

-l

TABLE 4-7 R VALUES - COAT'S HILK - CHILD m 2-sces/yr uC1/sec ,

NUCLIDE T. BODY .Cl-TRACT BONE LIVER KIDhEY THYROID LUNC H 3* 3.23E 03 3.23E 03 3.23E 03 3.23E 03 3.23E 03 3.23E 03 Cr 51 5.65E 03 3.00E 05 8.57E 02 3.14E 03 5.73E 03 Mn 54 3.44E 05 1.08E 06 1.29E 06 3.62E 05 Fe 59 5.88E 05 1.23E 06 7.29E 05 1.18E 06 3.41E 05, Co 58 2.12E 06 4.04E 06 6.92E 05 Co 60 8.17E 06 1.53E 07 2.77E 06 Zn 65 4.92E 08 1.39E 08 2.97E 08 7.91E 08 4.98E 08 L

Sr 89 1.87E 08 2.54E 08 6.56E 09 Sr 90 3.43E 10 1.82E 09 1.35E 11 Zr 95 4.27E 01 5.01E 04 2.18E 02 4.80E 01 6.57E 01 ,

! I 131 4.39E 08 6.88E 07 7.68E 08 7.72E 08 1.27E 09 2.55E 11 f

'i i I 133 4.93E 06 5.25E 06 1.05E 07 1.30E 07 2.17E 07 2.42E 09 Cs 134 1.23E 10 3.14E 08 3.55E 10 5.82E 10 1.80E 10 6.47E 09 }

Cs 136 2.56E 09 1.39E 08 1.44E 09 3.96E 09 2.11E 09 3.14E 08 Cs 137 7.57E 09 3.21E 08 5.36E 10 5.13E 10 1.67E 10 6.01E 09 ,

Ba 140 3.92E 05 3.41E 06 6.72E 06 '5.89E 03 1.92E 03 3.51E 03 1

Ce 141 8.97E 01 7.54E 05 1.21E 03 6.04E 02 2.65E 02  !

  • mrem /yr uCi/mJ

TABLE 4-8 R VALUES - COAT'S MILK - TEEN m2-erea/y r uC1/sec ,

i NUCLIDE T. BODY Cl-TRACT BONE LIVER KIDNEY 1hYROIL LUNG I H 3* 2.04E 03 2.04E 03 2.04E 03 2.04E 03 2.04E 03 2.,04E 03

, Cr 51 2.77E 03 4.66E 05 6.07E 02 1.54E 03 3.95E 03 Hn 54 1.71E 05 1.77E 06 8.64E 05 2.58E 05 f Fe 59 2.83E 05 1.74E 06 3.14E 05 7.34E 05 I 1.31E 05 .

i Co 58 1.04E 06 6.25E 06 4.63E 05 Co 60 4.02E 06 2.32E 07 1.78E 06 2n 65 2.45E 08 2.22E 08 1.51E 08 5.25E 08 3.36E 08 {

Sr 89 7.59E 07 3.16E 06 2.65E 09 Sr 90 1.98E 10 2.25E 09 8.81E 10 Zr 95 2.04E 01 -6.84E 04 9.40E 01 2.97E 01 4.36E 01

~

1 131 2.38E 08 8.77E 07 3.17E 08 4.43E 08 7.63E 08 1.29E 11 I 133 2.24E 06 -

5.57E 06 4.34E 06' 7.36E 06 1.29E 07 1.03E 09 Cs 134 1.68E 10 4.50E 08 1.54E 10 3.62E 10 1.15E 10 4.39E 09 Cs 136 1.69E 09 2.02E 08 6.38E 08 2.51E 09 1.37E 09 1.15E 08 Cs 137 1.03E 10 4.21E 08 2.22E 10 2.96E 10 1.01E 10 3.91E 09 Ba 140 1.80E 05 4.30E 06 2.79E 06 3.41E 03 1.16E 03 2.30E 03

. Ce 141 3.77E 01 9.39E 05 4.92E 02 3.28E 02 1.55E 02 i

  • mren/yr uC1/mJ

e TABLE 4-9

, R VALUES - COAT'S MILK - ADULT m2 -ares /yr uC1/sec I

NUCLIDE T. BODY Cl-TRACT BONE LIVER KIDhEY 1HYk0lb Lbh6 i

X 3* 1.57E 03 1.57E 03 1.57E 03 1.57E 03 1.57E 03 1.57E 03

  • Cr 51 1.59E 03 3.99E 05 3.49E 02 9.48E 02 2.11E 03 Mn 54 9.89E 04 1.59E 06 5.19E 05 1.54E 05 i i

Fe 59 1.62E 05 1.41E 06 1.80E 05 4.23E 05 1.16E 05 Co 58 6.03E 05 5.46E 06 2.69E 05 4

i Co 60 2.32E 06 1.98E 07 1.0$E 06 i Zn 65 1.42E 08 1.97E 08 9.85E 07 3.14E 08 2.10E 08 l

Sr 89 4.13E 07 2.31E 08 1.44E 09 Sr 90 1.39E 10 1.64E 09 5.67E 10 Zr 95 1.17E 01 5.46E 04 5.37E 01 1.72E 01 2.70E 01 i

1 131 1.43E 08 6.59E 07 1.74E 08 2.50E 08 4.28E 08 8.18E 10 '

1 133 1.26E 06 3.71E 06 2.37E 06 4.13E 06 7.21E 06 6.07E 08  ! !

Ca 134 1.72E 10 3.69E 08 8.85E 09 2.11E 10 6.82E 09 2.26E 09 l-Cs 136 1.06E 09 1.68E 08 3.75E 08 '1.48E 09 8.23E 08 1.13E 06 I

, Cs 137 1.10E 10 3.25E 08 1.25E 10 1.68E 10 5.70E 09 1.89E 09 l Ba 140 1.01E 05 3.18E 06 1.54E 06 1.94E 03 6.59E 02 1.11E 03 ,

Ce 141 2.06E 01 6.94E 05 2.68E 02 1.81E 02 8.43E 01 l

  • mrea/yr uC1/m' s ,.

. TABLE 4-10 R VALUES - MEAT - CHILD m2 -erea/yr uC1/se c a

NUCLIDE T. BOD! Cl-1RACT B0hh LIVER KILhEi 1hYkolb LbhG H 3* 2.36E 02 2.36E 02 2.36E 02 2.36E 02 2.36E 02 2.36E 02 Cr 51 4.07E 03 2 .16E 05 6.17E 02 2.26E 03 4.12E 03 Mn 54 1.09E 06 3.45E 06 4.11E 06 1.15E 06 i

Fe 59 1.42E 08 2.97E 08 1.76E 08 2.85E 08 8.26E 07  :

I Co 58 2.39E 07 4.56E 07 7.82E 06  ;

Co 60 1.09E 08 2.05E 08 3.70E 07 Zn 65

  • 3.72E 08 1.0$E 08 2.25E 08 5.99E 08 3.77E 08 Sr 89 6.55E 06 8.87E 06 2.29E 08  ;

i Sr 90 1.52E 09 8.08E 07 6.00E 09  !

Zr 95 2.48E 05 2.91E 08 1.27E 06 2.79E 05 3.99E 05 1 131 4.64E 06 7.29E 05 8.14E 06 8.19E 06 1.34E 07 2.71E 09 1 133 1.55E-01 1.66E-01 3.32E-01 4.11E-01 6.65E-01 7.63E 01 i

Cs 134 1.67E 08' 4.26E 06 4.81E 08 7.90E 08 2.45E 08 6.7bE 07 Cs 136 1.35E 07 7.34E 05 7.60E 06 2.09E 07 1.11E 07 1.66E 06 Cs 137 1.04E 08 4.43E 06 7.39E 08 7.07E 08 2.30E 08 8.29E 07 Ba 140 1.22E 06 1.06E 07 2.10E 07 1.84E 04 5.96E 03 1.10E 04 .

I Ce 141 7.57E 02 6.36E 06 1.02E 04 5.10E 03 2.24E 03

  • mres/yr uC1/mJ '

a .

., TABLE 4-11 R VALUES - MEAT - TEEN m2-aren/yr

.uci/se c

'NUCLIDE T. BODY Cl-TRACI B0hh LIVER KILhEi 1hikO1L lbh(. '

i-11 3* 1.95E 02 1.95E 02 1.95E 02 1.95E 02 1.95E 02 1.95E Di Cr 51 2.61E 03 4.39E 05 5.72E 02 ~1.45E 03 3.73E 03 g Hn 54 7.12E 05 7.37E 06 3.59E 06 1.07E 06 ,

Fe 59 8.95E 07 5.48E 08 9.93E 07 2.32 E 08 - 7.31E 07, Co 58 1.54E 07 9.22E 07 6.69E 06 '

Co 60 7.03E 07 4.06E 08 3.12E 07 Zn 65 2.43E 08 2.20E 08 1.50E 08 5.20E 08 3.33E 08 Sr 89 3.47E 06 1.44E 07 1.21E 08 Sr 90 1.15E 09 1.30E 08 4.64E 09 l

Zr 95 1.55E 05 5.20E 08 7.15E 05 2.25E 05 3.31E 05 j 1 131 3.30E 06 1.22E 06 4.39E 06 6.14E 06 1.06E 07 1.79E 09 4

1 133 9.25E-02 2.30E-01 1.79E-01 3.03E-01 5.32E-01 4.13E 01 i

Ca 134 2.98E 08 7.99E 06 2.73E 08 6.42E 08 2.04E 08 7.76E 07 Ca 136 1.16E 07 1.40E 06 4.41E 06 1.73E 07 9.44E 06 1.49E 06 i Cs 137 1.86E 08 7.59E 06 4.01E 08 .5.34E 08 1.82E 06 7.06E 07

\

Ba 140 7.33E 05 1.75E 07 1.14E 07 1.39E 04 4.72E 03 9.37E'03-Ce 141 4.14E* 02 1.04E 07 5.43E 03 3.63E 03 1.71E 03 f

  • ares /yr uC1/m3

1 TABLE 4-12 '

R VALUES - MEAT - ADULT m2 -area /yr uC1/sec t ,

NUCLIDE T. BODY Cl-IRAC1 BOhE LIMA K1bhElf ih'rk0lb Lbhb i

H 3* 3.27E 02 3.27E 02 3.27E 02 3.27E 02 3.27E O2 3.27E 01 Cr 51 3.26E 03 8.21E 05 7.19E 02 1.95E 03 4.33E 03 i

! i 1 Mn 54 8.98E 05 1.44E 07 4.71E 06 1.40E 06 ,

Fe 59 k 1.12E 08 9.73E 08 1.24E 08 2.92E 08 6.16E 07 i Co 58 1.95E 07 1.76E 08 8.68E 06 Co 60 8.87E 07 7.56E,08 4.02E 07 4

. Zn 65 3.06E 08 4.27E 08 2.13E 08 6.78E 08 4.53E 08 Sr 89 4.12E 06 2.30E 07 1.43E 08 l

Sr 90 1.76E 09 2.07E 08 7.17E 09 Zr 95 1.94E 05 9.07E 08 8.92E 05 2.86E 05 4.49E 05 I

I 131 4.33E 06 1.99E 06 5.28E 06 7.55E 06 1.29E 07 2.48E 09 1 133 1.13E-01 3.34E-01 2.14E-01 3.72E-01 6.49E-01 5.46E 01 Cs 134 6.68E 08 1.43E 07 3.43E 08 8.15E 08 2.64E 06 6.76E 07 i

Cs 136 1.61E 07 2.53E 06 5.65E 06 2.23E 07 1.24E 07 1.70E b6 i ca 137 4.33E 08 1.28E 07 4.83E 08 6.61E 08 2.24h 06 7.46E 07

Ba 140 9.01E 05 2.83E 07 1.38E 07 1.73E 04 5.87E 03 9.69E D3 j I

Ce 141 4.96E C2 1.67E 07 6.47E 03 4.38E 03 2.03E 03 l, i

  • ares /yr I j uC1/mJ

)

TABLE 4-13 R VALUES - VECETATION - CHILD mbrea/yr uC1/se c NUCLIDE T. BODY GI-1RACT ,y LIVER KIDhEY 1hYkOIL LUNG 113* 4.04E 03 4.04E 03 4.04E 03 4.04E 03 4.04E 03 4.04E 03 4

. Cr 51 1.16E 05 6.15E 06 1.76E 04 6.44E 04 1.16E 05 Hn 54 1.73E 08 5.44E 08 6.49E 08 1.82E 06 Fe 59 3.17E 08 6.62E 08 3.93E 06 6.36E 08 '

1.64E 08, Co 58 1.92E 08 3.66E 08 6.27E 07 -

F Co 60 1.11E 09 2.08E 09 3.76E'08 l 1

Zn 65 1.70E 09 4.81E 08 1.03E 09 2.74E 09 1.73E 09 Sr 89 1.03E 09 1.40E 09 3.62E 10 l-Sr 90 3.49E 11 1.86E 10 1.38E 12  !-

Zr 95 7.44E 05 8.71E 08 3.80E 06 8.35E 05 1.20E 06 1 131 8.16E 07 1.28E 07 1.43E 08 1.44E 08 2.36E 08 4.75E 10 1 133 1.67E 06 1.78E 06 3.57E 06 4.42E'06 7.36EN Mf q r821E 08 Cs 134 5.40E 09 1.38E 08 1.56E 10 2.56E 10 7.93E 09

' ' o c?,d4 E 0 9 Cs 136

, 1.43E 08 7.77E 06 8.04E 07 2.21E 06 1.18E 06 1.7bh 07% ,

't I

Cs 137 3.52E 09 1.50E 08 2.48E 10 2.39E 10 7.78E 09 2.60E 09 '\ };

j Ba 140 1.61E 07 1.40E 08 2.76E 08 2.42E 05 7.67E 04 1.44E 05 Ce 141 4.75E 04 3.99E 08 6.42E 05 3.20E 05 1.40E 05

  • nren/yr uC1/mJ i'

1

TABLE 4-14 R VALUES - VECETATION - TEEN

! a b ren/yr uC1/5tc i

, NUCLIDE T. BODY Cl-TRACT BONE LIMR KIDhEY 1hYR01D LUE

! H 3* 2.61E 03 2.61E 03 2.61E 03 2.61E 03 2.61E 03 2.61E 03 l

Cr 51 6.11E 04 1.03E 07 1.34E 04 3.39E 04' 8.72E 04 i

4 Mn 54 8.79E 07 9.89E 08 4.43E 08 1.32E 08 i {-

Fe 59 1.60E 08 9.78E 08 1.77E 06 .4.14L 06 1.3bE 08, Co 58 9.79E 07 5.85E 08 4.25E 07 i

Co 60 5.57E 08 3.22E 09 2.47E 08 j Zn 65 8.68E 08 7.88E 08 5.36E 08 1.86E 09 1.19E 09 Sr 89 4.36E 08 1.81E 09 1.52E 10 Sr 90 2.05E 11 2.33E 10 8.32E 11 f Zr 95 3.68E 05 1.23E 09 1.69E 06 5.35E 05 7.86E 05 .

V

] 1 131 5.77E 07 2.13E 07 7.68E 07 1.07E 08 1.85E 08 3.14E 10 g j 1 133 1.01E 06 2.51E 06 1.96E 06 3.32E 06 5.83E 06 4.64E 08 i Cs 134 7.54E 09 - 2.02E 08 6.90E 09 1.62E 10 5.16E 09 1.97E b9 '

l j Cs 136 1.13E 08 1.35E 07 4.28E 07 1.68E 08 9.16E 07 1.44E 07

Cs 137 4.90E 09 2.00E 08 1.06E 10 1.41E 10 4.78E 09 1.66E 09 1i l Ba 140 8.88E 06 2.12E 08 1.38E 06 1.69E 05- 5.72E 04 1.14E 05

, Ce 141 2.12E 04 5.29E 08 2.77E 05 1.85E 05 8.70E 04 '

1

  • area /yr 4

uC1/mJ

. . 'C TABLE 4-15 R VALUES - VEGETATION - ADULT m 2-ares /yr uC1/sec NUCLIDE T. BODY Cl-TRACT BONE LIVER KIbNEY 1hYROIL LUNG

! H 3* 2.28E 03 2.28E 03 2.28E 03 2.28E 03 2.28E D3 2.28E 03 Cr 51 4.60E 04 1.16E 07 1.01E 04 2.75E 04 6.lbE 04

! Mn 54 5.83E 07 9.36E 08 3.05E 08 9.09E 07 Fe 59 1.12E 08 9.75E 08 1.24E.08 2.93E 06 6.17E 07, Co 58 6.71E 07 6.07E 08 2.99E 07 Co 60 3.67E 08 3.12E 09 1.6bE 06 i

Zn 65 5.77E 08 8.04E 08 4.01E 08 1.28E 09 8.54E 08 i t

i Sr 89 2.87E 08 1.60E 09 1.08E 10 Sr 90 1.64E 11 1.93E 10 6.70E 11 Zr 95 2.51E 05 1.17E 09 1.16E 06 3.71E 05 5.82E 05 1 131 6.61E 07 3.04E 07 8.07E 07 1.15E 08 1.98E 08 3.78E 10 j

1 133 1.12E 06 3.30E 06 2.11E 06 3.67E 06 6.40E 06 5.39E 08 Ca 134 8.83E 09 1.89E 08 4.54E 09 1.08E 10 3.49E 09 1.16E 09 Cs 136 1.19E 08 1.88E 07 4.19E 07 1.66E 08 9.21E 07 1.26E 07 Cs 137 5.94E 09 1.76E 08 6.63E 09 9.07E 09 3.08E 09 1.02E 09 Ba 140 8.40E 06 2.64E 08 1.28E 08 1.61E 05 5.47E 04 9.22E 04 Ce 141 1.48E 04 4.99E 08 1.93E 05 1.31E 05 6.07E 04

[

]

  • area /yr h

uC1/a J

~

TABLE 4-16 R VALUES . INHALATION - INFANT area /yr

[ uC1/m3 NUCLIDE T. BODY Cl-TRAC 1 BONE LIVER KIDNEY 1HYR01D LUNG H3 6.46E 02 6.46E 02 6.46E 02 6.46E 02 6.46E 02 6.46E 02 i Cr 51 8.93E 01 3.56E 02= 1.32E 01 5.75E 01 1.28E 04 Hn 54 4.98E 03 7.05E 03 2.53E 04 4.98E 03 9.98E 05 Fe 59 9.46E 03 0.47E 04 1.35E 04 2.36E 04 '

1.01E 06, Co 58 1.82E 03 1.11E 04 1.22E 03 7.76E 05 Co 60 1.18E 04 3.18E 04 6.01E 03 j 4.5bh 06 2n 65 3.10E 04 5.13E 04 1.93E 04 6.25E 04 3.24h 04 6.46L 05 t Sr 89

(

1.14E 04 6,39E 04 3.97E 05 2.05E 06 I,

Sr 90 2.59E 06 1.31E 05 4.08E 07 1.12E 07 I

l Zr 95 2.03E 04 2.17E 04 1.15E 05 2.78E 04 3.10E 04 1.75E 06 Ii l

l 1 131 1.96E 04 1.06E 03 3.79E 04 4.43E 04 5.17E 04 1.48E 07 I 133 5.59E 03 2.15E 03 1.2 E 04 1.92E 04 2.24E 04 3.55E 06

's Ca 134 7.44E 04 1.33E 03 3.96E 05 7.02E 05 1.90E 05 7.95E 04  ;

i:

Ca 136 5.28E 04 1.43E 03 4.82E 04 1.34E 05 5.63E 04 1.17E 04 I l Cs 137 4.54E 04 1.33E 03 5.48E 05 6.31E 05 1.72E 05 7.12E 04 Ba 140 2.89E 03 .3.83E 04 5.59E 04 5.59E 01 1.34E 01 1.59E 06

~

! Ce 141 1.99E 03 2.15E 04 2.77E 04 1.66E 04 5.24E 03 5.16E 05 f

o. .

TABLE 4-17 l R VALUES - INHAIATION - CHILD

, mres/yr uC1/m 3 NUCLIDE T. BODY Cl-IRACT BONE LIVIJt KIDhEY 1hYR01D LUNG

H3 1.12E 03 1.12E 03 1.12E 03 1.12E 03 1.12E 03 1.12E 03

! Cr 51 1.54E' 02 1.08E 03 2.43E 01 8.53E 01 1.70E 04 i l

Mn 54 9.50E 03 . 2.29E 04 4.29E 04 1.00E 04 1.57E 06 Fe 59 1.67E 04 7.06E 04 2.07E 04 3.34E 04 1.27E 06 ,

Co 58 3.16E 03 3.43E 04 1.77E 03 1.10E 06

)

Co 60 2.26E 04 9.61E 04 1.31E 04 7.0th b6 2n 65 7.02E 04 1.63E 04 4.25E 04 1.13E 05 7.13E 04 9.94E 05 Sr 89 1.72E 04 1.67E 05 5.99E 05 1.1SE b6 Sr 90 6.43E 06 3.43E 05 1.01E 08 1.47E 07 '

Zr 95 3.69E 04 6.10E 04 1.90E 05 4.17E 04 5.95E 04 2.23E D6 1

I 131 2.72E 04 2.84E 03 4.80E 04 4.80E 04 7.87E 04 1.62E 07

1 133 7.68E 03 5.47E 03 1.66E 04 2.D3E 04 3.37E 04 3.64E 06 Cs 134 2.24E 05 3.84E 03 6.50E 05 1.01E 06 3.30E 05 1.21E 05 Ca 136 1 16E 05 4.17E 03 6.50E 04 1.71E 05 9.53E 04 1.45E 04 Cs 137 1.28E 05 3.61E 03 9.05E 05 8.24E 05 2.82E 05 1.04E 05 ,,

!i' Ba 140 4.32 E 03 1.02E 05 7.39E 04 6.47E 01 2.11E 01 1.74E 06 Ce 141 2.89E 03 5.65E 04 3.92E 04 1.95E 04 8.53E 03 5.43E 05 i .

MJ -

\

s TABLE 4-18 R VALUES - INHALATION - TEEN aren/yr l uC1/m 3 NUCLIDE T. BODY CI-TRACT BONE LIVER K1DhhY 1hYRO1D Lbhb 11 3 1.27E 03 1.27E 03 1.27E 03 1.27E 03 1.27E 03 1.27E 03 Cr 51 1.35E 02 3.00E 03 3.07E 01 7.49E 01 2.09E 04 Mn 54 8.39E 03 6.67E 04 5.10E 04 1.27E 04 1.98E 06 Fe 59 1.43E 04 1.78E 05 1.59E 04 3.69E 04 1.53E 06, Co 58 2.77E 03 9.51E 04 2.07E 03 1.34E 06 Co 60 1.98E 04 2.59E'05 1.51E 04 6.71E 06 Zn 65 6.23E 04 4.66E 04 3.85E 04 1.33E 05 8.63E 04 1.24E 06 l l Sr 89 1.25E 04 3.71E 05 4.34E 05 2.41E 06 1  :

Sr 90 6.67E 06 7.64E 05 1.08E 08 1.bSE 07 l Zr 95 3.15E 04 1.49E 05 1.45E 05 4.58E 04 6.73E 04 1.6bE 06 1 131 2.64E 04 6.48E 03 3.54E 04 4.90E 04 8.39E 04 1.46E 07 1 133 6.21E 03 1.03E 04 1.21E 04 2.05E 04 3.59E 04 2.92E 06 Cs 134 5.48E 05 9.75E 03 5.02E Os 1.13E 06 3.75E 05 1.4bE DS Cs 136 1.37E 05 1.09E 04 5.14E 04 1.93E 05 1.10E 05 1.??E 04 Cs 137 3.11E 05 8.47E 03 6.69E 05 8.47E 05 3.04E 05 1.21E 05 Ba 140 3.51E 03 2.28E 05 5.46E 04 6.69E 01 2.24E 01 2.03E 06 I ce 141 2.16E 03 1.26E 05 2.84E 04 1.89E 04 8.8'E 03 6.13E 05 i

TABLE 4-19 R VALUES - INHAIATION - ADOLT ares /yr uC1/m 3 i

! NUCLIDE T. BODY Cl-TRACT BONE L1VER K1bhhi 1hYk0lb LbhG H3 1.26E 03 1.26E 03 1.26E 03 1.26E 03 1.26E D3 1.26E 03 i Cr 51 9.99E 01 3.32E 03 2.28E 01 ' 5.94E 01 1.44E 04 t1

Mn 54 6.29E 03 7.72E 04 3.95E 04 9.83E 03 1.40E 06 i Fe 59 1.05E 04 1.88E 05 1.17E 04 2.77E 04 1.01E 06  !

Co 58 2.07E 03 1.06E 05 1.58E 03 9.27E 05 j 1

Co 60 1.48E 04 2.84E 05 1.15E 04 5.96E 06 i Zn 65 4.65E 04 5.34E 04 3.24E 04 ~1.03E 05 6.89E 04 8.63E 05 l j Sr 89 8.71E 03 3.49E 05 3.04E 05 1.4bE D6 l

! Sr 90 6.09E 06 7.21E 05 9.91E 07 9.59E 06 l

! 2.32E 04 Zr 95 1.50E 05 1.07E 05 3.44E 04 5.41E 04 1.7?E 06 l

l 1 131 2.05E 04 6.27E 03 2.52E 04 3.57E 04 6.12E 04 1.19E 07 '

i 1 133 4.51E 03 8.87E 03 6.63E 03 1.4BE 04 1.56E 04 2.15E 06 Ca 134 7.27E 05 1.04E 04 3.72E 05 8.47E 05 2.87E 05 9.75E'04 i Cs 136 1.10E 04 1.17E 04 3.90E 04 1.46E 05 6.55E 04 1.20E 04 l

j Cs 137 4.27E 05 8.39E 03 4.78E 05 6.20E 05 2.22E 05 7.51E 04 Ba 140 2.56E 03 2.18E 05 3.90E 04 4.90E 01 1.67E 01 1.2?E 06 Ce 141 1.53E 03 1.20E 05 1.99E 04 1.35E 04 6.25E 03 3.61E 05 l .

, . ._ . _ . _ _ _ . _ . _ _ _ _ _ . . ~ . . . . _ _ _ _ . _

2 TABLE 4-20 R VALUES - GRQUND - ALL AGE GROUPS m 4 ram /vr uCi/sec NUCLIDE T. BODY ,

Cr 51 4.66E 06 Mn 54 1.34E 09 Fe 59 2.75E 08 Co 58 3.79E 08 Co 60 2.15E 10 Zn 65 7.49E 08 Sr 89 2.23E 04 Zr 95 2.49E 08 I 131 1.72E 07 I 133 2.47E 06 Cs 134 6.82E 09 Cs 136 1.49E 08 Cs 137 1.03E 10 Ba 140 2.0$E 07 Ce 141 1.36E 07 l I 135 Later i

Ba/La*140 Later Nb 95 I4ter Sb 125 Later l

I

. - - ~ . - - . .-- . _. __ - - . . -

.l l

o .)

Nine Mile Roint Nuclear Station Unit 1

. Radiological Environmental' Monitoring Program

. Sampling Locations 1

hble 5.1 i

Type of

  • Map .

( ky .

Sample Iccation Collection Sice Frogram No.) Location Radiciodine and 1 Nine Mile Point Road 1.8 mi G 88* E Particulates (air) north (R-1)

Radioiodine and 2 Co. Rt. 29 & Lake Road (R-2) 1.1 mi 6 104* ESE Particulates (air)

Radiciodine and 3 Co. Rt. 29 (R-3) 1.5 mi G 132' SE i Particulates (air)

Radioiodine and 4 Village of Lycoming, NY (R-4) 1.8 mi 4 143' SE Particulates (air)

Radiciodine and 5 Montario Point Road (R-5) 16.4 mi 6 42' NE l Particulates (air) l I

j Direct Radiation (TLD) 6 North Shoreline Area (75) 0.1 mi 8 5' N j Direct Radiation (TLD) 7 North Shoreline Area (76) 0.1 mi 6 25' NNE  !

Direct Radiation (TLD) 8 North Shoreline Area (77) 0.2 mi 8 45' NE i i

Direct Radiation (TLD) 9 North Shoreline Area (23) 0.8 mi 9 70' ENE l Direct Radiation (TLD) 10 JAF east boundary (78) 1.0 mi G 90* E Direct Radiation (TLD) 11 Rt. 29 (79) 1.1 mi 8 115' ES E Direct Radiation (TLD) 12 Rt. 29 (80) 1.4 mi 8 133' SE Direct Radiation (TLD) 13 Miner Road (81) 1.6 mi 8159' SSE Direct Radiation (TLD) 14 Miner Road (82) 1.6 mi 6 181* S l Direct Radiation (TLD) 15 lakeview Road (83) 1.2 mi 6 200* SSW t

Direct Radiation (TLD) 16 l Lakeview Road (84) 1.1 mi 6 225' SW Direct Radiation (TLD) 17 Site Meteorological Ibwer (7) 0.7 mi 6 250' WSW Direct Radiation (TLD) 18 Enerav A formation Center (18)- 0.4 mi 9 255' W l

CMap - See Figures 5.1-1 and 5.1-2

~

i

_ _ . _. _ _ _ . _ . _ _ _ _ _ _ _ _ ____.- -1 o

Nine Mile Point Nuclear Station Unit 1 Radiological Environmental Monitoring Program

,. Sampling Iocations

. Table 5.1 (Continued)

Type of

  • Map (Env. )

Sample Iccation Collection Site (Program No.) Location Direct Radiation (TLD) 19 North Shoreline (85) 0.2 mi G 294' WNW 2 Direct Radiation (TLD) 20 North Shoreline (86) 0.1 mi 6 315' NW Direct Radiation (TLD) 21 North Shoreline (87) 0.1 mi 9 341' NNW Direct Radiation (TLD) 22 Hickory Grove (88) 4.5 mi G 97 ' E Direct Radiation (TLD) 23 Leavitt Road (89) 4.1 mi 8 lil' ESE Direct Radiation (TLD) 24 Rt. 104 (90) 4.2 mi 9135' SE Direct Radiation (TLD) 25 Rt. 51A (91) 4.8 mi 6 156' SSE Direct Radiation (TLD) 26 Maiden Iane Road (90) 4.4 mi 8 183* S ,

. I Direct Radiation (TLD) 27 Co. Rt. 53 (93) 4.4 mi 8 205' SSW j i

Direct Radiation (TLD) 28 Co. Rt.1 (94) 4.7 mi 6 223' SW j Direct Radiation (TLD) 29 Lake Shoreline (95) 4.1 mi 6 237' WSW  !

Direct Radiation (TLD) 30 Phoenir, NY Control (49) 19.8 mi 8 170' S l

Direct Radiation (TLD) 31 S.W. Oswego, Control (14) 12.6 mi 6 226' SW  !

Direct Radiation (TLD) 32 Scriba, NY (96) 3.6 mi 6199' SSW Direct Radiation (TLD) 33 Alcan Aluminum, Rt. lA (58) 3.1 mi 8 220' SW Direct Radiation (TLD) 34 Lycoming, NY (97) 1.8 mi G 143' SE Direct Radiation (TLD) 35 New Haven, NY,(56) 5.3 mi 8123* ESE Direct Radiation (TLD) 36 W. Boundary, Bible Carp (15) 0.9 mi 9 237' WSW Direct Radiation (TLD) 37 Lake Road (98) 1.2 mi 6 101' E Surface Water 38 OSS Inlet Canal (NA) 7.6 mi G 235' SW Surface Water 39 JAFNPP Inlet Canal (NA) 0.5 mi 8 70* ENE (NA) = not applicable

  • Map - See Figures 5.1-1 and 5.1-2 '

______1_

, . . . . - . -. - . . - - . .- .. .- . . - ~ . . - . . - - .

V ,

J

, Nine Mile Point Nuclear Station thit 1 Radiological Environmental Monito::ing Program Sampling Incations I

Table 5.1 l (Continued)

Type of

  • Map' ( hv. ) '

Sample Incation Collection Site (Program No.) Location Shoreline Sediment 40 Sunset Bay Shoreli u (NA)~ 1.5 mi 8 80' E 2

Fish 41 NMP Site Discharge Area (NA) 0.3 mi- t 315' NW and/or Fish 42 NMP Site Discharge Area (NA) 0.6 mi 4 55' NE Fish 43 Oswego Harbor Area (NA) 6.2 mi 9 235' SW Milk 44 Milk Iocation #50 (NA) 9.3 mi 6 93* E Milk 45 Milk Location'#7 (NA) 5.5 mi 8 107' ESE I

Milk 46 Milk Incation #16 (NA) 5.9 mi e 190* S  !

Milk 47 Milk Location #40 (NA) 15.0 mi 6 223* SW Food Product 48 Produce location #6** 1.9 mi 6 143' S E  !

(Bergenstock) (NA)  ;

I Food Product 49 Produce Incation #1** 1.8 mi 6 96* E i

' (J. Parkhurst) (NA)  !

i Food Product 50 Produce Incation #2** 1.9 mi 6101' E .I (Fox) (NA) '

l Food Product 51 Produce Iccation #5** 1.5 mi 0 114' ES E (C.S. Parkhurst) (NA)

Food Product 52 Produce Iccation #3** 2.3 mi 6 122* ES E l (T. Parkhurst) (NA) i l Food Product 53 Produce Incation #4** 2.2 m'i e 123' ES E l (C. Lawton) (NA)  !

l I'ood Product 54 Produce Incation #7** 15.0 mi 6 223' SW j (Mc Millen) (NA)

Food Product 55 Produce Iocation #8** 12.6 mi 6 225' SW (Denman) (NA)

  • Map - See Figures 5.1-1 and 5.1-2
    • Food Product samples need not necessarily be collected from all listed locations. Collected samples will be of the highest calculated site average D/Q.

(NA) = nott applicable I

_ _ __ . _ _ _ _ . . _ _ . _ . - -__ - = . , _ _ _ . _ _ _ _ _ _____ __

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