ML19347A470

Offsite Dose Calculation Manual

ML19347A470
Person / Time
Site:	Yankee Rowe
Issue date:	04/09/1979
From:	YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML19347A467	List: ML19347A466 ML19347A470
References
PROC-790409, NUDOCS 7904160197
Download: ML19347A470 (58)

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OFFSITE DOSE CALCULATION MANUAL 50-2.9 i

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TABLE OF CONTENTS Section PR 1

1.0 INTRODUCTION

3 2.0 LIQUID RELEASE CLSE CALCULATIONS 3

2.1 Technical Specification 3.11.1.2, Dose to an Individual 4

2.1.1 Method I 4

2.1.1.1 Dose to the Total Body 4

2.1.1.2 Dose to the Critical Organ 4

2.1.1.3 Application of Method I 5

2.1.2 Method II f

2.1.2.1 Dose to the Total Body 2.1.2.2 Dose to the Critical Organ 6

2zl.2.3 Application of Method II 6

2.1.3 Method III 7

3.0 GASEOUS RELEASE DOSE CALCULATIONS 7

3.1 Technical Specification 3.11.2.1, Dose Rate Limits 7

3.1.1.

Method I 7

3.1.1.1 Dose Rate Due to Noble Gases' 7

3.1.1.1.1 Dose Rate to the Total Body 8

3.1.1.1.2 Dose Rate to the Skin 8

3.1.1.2 Dose Rate to the Critical Organ Due to Radioiodines and Particulates O

3.1.1.3 Application of Method I 9

3.1.2 Method II 11

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TABLE OF CONTENTS (Continued)

Section P_ag 11 3.2 Technical Specifications 3.11.2.2, Dose to Air Due to Noble Gases 11 3.2.1 Method I 3.2.1.1 Air Dose Due to Gamma Radiation 11 11 3 2.1.2 Air Dose Due to Beta Radiation 12 3.2.1.3 Application of Method 1 12 3.2.2 Method II 14 3.3 Technical Specification 3.11.2.3, Dose to an Individual 14 3.3.1 Method I 15 3.3.1.1 Dose to the Thyroid 15 3.3.1.2 Application of Method I 15 3.3 2 Method II 16 4.0 METEOROLOGY 18 5.0 ENVIRONMENTAL MONITORING 23 APPENDIX A - Basis for the Dose Calculation Methods 40 APPENDIX B - Basis for the Atmospheric Dilution Factors 43 APPENDIX C - Basis for Setpoint Determinations 51 APPENDIX D - Radwaste Equipment

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1.0 INTRODUCTION

The purpose of this manual is to provide methods to insure compliance with the dose requirements of the Technical Specifications. Each method is based on a plant specific application of the models presented in Regulatory Guide 1.109.(1)

Methods are included to calculate the doses to individuals from both gaseous and liquid releases from the plant.

Under normal operations, experience has shown that the plant will be operated at a small fraction of the dose limits imposed by the Technical Specifications. For this reason the dose evaluations are presented at different levels of sophistication. The first method being the most conservative, but simplest to use, subsequent methods somewhat more entailed and more realistic while the final method requires a full analysis following the guidance presented in Regulatory Guide 1.109.

The first method is based on a critical organ, critical age group and as such it provides a conservative estimate of the doses required by the Technical Specifications.

If the Technical Specifications are met by application of the first method, no further analysis will be required.

If, however, Method One indicates that the Technical Specification limits are being approached, a more realistic estimate may be obtained by application of subsequent methods.

l The final method will calculate the dose to seven organs of four age groups and is based on measured releases for each nuclide. This method

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will be used to assess doses for the Semi-Annual Effluent Report. _

The basis for each of the dose calculation methods is described

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in the Appendix A.

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2.0 LIQUID RELEASE DOSE CALCULATIONS 2.1 Technical Specification 3.11.1.2, Dose to an Individual This section is to be used to insure compliance with the following Technical Specification:

LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose commitment to an individual from radioactive materials in liquid effluents released to unrestricted areas shall be limited:

a.

During any calendar quarter to <1.5 mrem to the total body and to f,5 mrem to any organ, and b.

During any calendar year to I3 mrem to the total body and to f,10 mrem to any organ.

The dose commitment to any individual from liquid releases is proportional to the quantity (curies) to which that individual is exposed.

The following equations shall be used to calculate the dose commitment resulting from a liquid release (in curies) from the Yankee Rowe Station.

The specification requires a monthly evaluation, however the following equations can be applied for any duration of release.

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2.1.1 Method I 2.1.1.1 Dose to the Total Body The dose to the total body is:

Dtb(mrem) =

10 Q137Cs where:

Q137Cs = Cesium-137 Release (C1) 2.1.1.2 Dose to the Critical Organ The dose to the critical organ is:

Dorgan(mrem) = 20 Q137Cs + Q131I

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where:

131I.

= I dine-131 Release (Ci)

Q Q137Cs = Cesium-137 Release (Ci) l

~.1.1 3 Application of Method I 2

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Step 1.

Determine the number of curies of Cesium 137 and Iodins 131 released during the period.

Step 2.

Perform the above multiplications to obtain the total body and i

organ doses for the period.

1 I'

Step 3.

Record the~ total body and organ doses and maintain a cumulative k

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dose for the annual and quarterly periods.

2.1.2 Method II 2.1.2.1 Dose to the Total Body The dose to the total body is:

Dtb(mrem) = 0.015 089Sr + 2.3 0 0Sr + 3.9x10-6 Q3H 9

+ 10.0 Q134cs + 5.9 Q137cs where:

Q 0Sr =

The respective releases based on the last

Qggsp, 9

measurements available for Strontium-89 and Strontium-90(ci).

Q3H. Q134cs. 0137cs : The respective releases based on the present measurement for Tritium, cesium-134 and Cesium-137(ci).

2.1.2.2 Dose to the Critical Organ 89Sr + 9.4 0 0Sr + 3.9x10-6 Q3H + 1.1 Q1311 D

l organ (mrem) = 0.53 0 9

+ 12.4 Q134cs + 9.4 Q137cs where:

i Q89Sr.Q 0Sr =

The respective releases tused on the last l

9 measurements available for Strontium-89 and Strontium-90 (Ci).

k 0131I Q134cs, =

The respective releases based on the present Q137cs. 03H measurement for Iodine-131. Cesium-134. Tritium and Cesium-137 (Ci).

2.1.2.3 Application of Method II Step 1.

Determine the number of curies of Tritium, Cesium 134, Cesium 137. Tritium and Iodine 131 released during the period, i

Step 2.

Based on the last available measurements of Strontium-89 and Strontium-90; estimate the number of curies of each released during the present period.

Step 3 Obtain the Total Body and Organ doses by performing the above l

calculations.

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Step 4 Record the total body and organ doses and maintain a cumulative dose for the quarterly and annual periods.

2.1.3 Method III The dose calculated shall be in conformance with Regulatory Guide 1.109 using site specific parameters applicable during the period of release.

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3.0 CASEOUS RELEASE DOSE CALCULATIONS 3.1 Technical Specification 3.11.2.1. Dose Rate Limits This section is to be used to insure compliance with the following Technical Specification:

0 LIMITING CONDITION FOR OPERATION 3.11.2.1 The dose rate in unrestricted areas due to radioactive materials released in gaseous effluents from the site shall be limited to the following:

The dose rate limit for noble gases shall be 1500 mrem /yr a.

to the total body, and 13000 mrem /yr to the skin b.

The dose rate limit for all radioiodines and radioactive materials in particulate form and radionuclides other than noble gases with half lives greater than 8 days shall be 11500 mrem /yr to any organ.

3.1.1 Method I 3.1.1.1 Dose Rate Due to Noble Gases 3.1.1.1.1 Dose Rate to the Total Body The total body dose rate due to noble gases can be determined as follows:

Dtb (8 '8) = 2.24 E Qi DFBi yr i

k

I where:

6g

= Release rate for each nuclide shown in Table 3.1 (uCi/sec).

DFBi = Total body dose factor (see Table 3.1) 3.1.1.1.2 Dose Rate to the Skin bakin ("yr }

• I Di DFi i

where:

b Release rate (uCi/sec) for each nuclide in Tables 3.1.

i =

i= Skin dose factor (see Table 3.1).

DF 3.1.1.2 Dose Rate to Critical Organ due to Radioiodines and Particulates i

The dose rate to the critical organ can be determined as follows:

organ I rem) = 3.2 x 103 Q3317 D

yr where:

b131I = Release rate for Iodine-131 (uCi/sec).

3.1.1.3 Application of Method I Step 1.

Determine the release rate in y Ci/sec (microcuries per second) for Iodine-131 and for each noble gas radionuclide detected.

Step 2.

Find the. dose rate to the total body by multiplying each noble gas release rate by the constant and by its total body dose factor

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4 determined from Table 31 and summing the con?.ributions from each nuclide.

Step 3 Find the gamma contribution to the skin dose by multiplying each noble gas release rate by its gamma air dose factor from Table 3 2.

Sum the contributions and multiply by the constant (2.5).

Similarly, to find the Beta contribution to skin dose, multiply each noble gas release rate by its skin dose factor from Table i

3.1.

Multiply by the constant (11) and add to the gamma contribution to obtain the total skin dose.

Step 4 Find the dose rate to the critical organ by multiplying the Iodine-17,1 release rate by 3.2 x 103 3.1.2 Method II f

The dose rates calculated will follow the guidance presented in Regulatory Guide 1.109 using site specific parameters applicable during the period i

of the release.

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Table 3.1 Noble Gases and Their Dose Factqrs (To be used for Technical Specification 3.11.2.1)

Total Body Dose Factor Skin Dose Factor

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DF' (mrem-sec) g ("pCi-yr F'"~"

DFB i

Ci-yr t

Kr-83m 7.56E-08' 4.83E-05 Kr-85m 1.17E-03 1.91E-02 Kr-85 1.61E-05 1.48E-02 Kr-87 5.92E-03 1.22E-01 Kr-88 1.47E-02 6.41E-02 Kr-89 1.66E-02 1.54E-01 Kr-90 1.56E-02 1.21E-01 Xe-131m 9.15E-05 5.63E-03 Xe-133m 2.51E-04 1.18E-02 Xe-133 2.94E-04 4.25E-03 Xe-135m 3.12E-03 1.62E-02 Xe-135 1.81E-03 2.53E-02 Xe-137 1.42E-03 1.38E-01 Xe-138 8.83E-03 6.85E-02 Ar-41 8.84E-03 5.28E-02

• 7.56E-08 = 7.56 x 10-8.

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32 Technical specification 3.11.?.2, Dose to Air Due to Noble Gase,s This section is to be used to insure compliance with the following Technical Specification:

LIMITING CONDITION FOR OPERATION 3 11.2.2 The air dose in unrestricted areas due to noble gases released in gaseous effluents shall be limited to the following:

a.

During any calendar quarter to 15 mrad for gamma radiation, and 110 mrad for beta radiation; and b.

During any calendar year,110 mrad for gamma radiation, and 120 mrad for beta radiation.

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3.2.1 Method I 3.2.1.1 Air Dose Due to Gamma Radiation Y

Y Dair (mrad) = 0.10 I Q DF g

g i

where:

Q

= Number of curies of nuclide "i" releasea.

g DF

= Gamma dose factor to air for nuclide "i". See Table 3.2.

g 3.2.1.2 Air Dose Due to Beta Radiation i

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Dair (arad) = 0.35 I Q DF g

g

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'where:

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= Beta dose factor to air for nuclide "i". See Table 3.2.

j g

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= Number of curies of nuclide "i" released.

1 3.2.1.3 Application of Methei I Step 1.

Determine the number of curles released during the period for each noble gas detected.

Step 2.

Find the dose to air due to gamma radiation by multiplying each noble gas released by the constant and by its gamma dose factor determined from Table 3.2.

Sum the contributions from each nuclide to find the total air dose from gamma radiation.

Step 3.

Find the dose to air due to beta radiation by multiplying each noble gas release by the constant and by its beta air dose factor i

determined from Table 3.2.

Sum the contributions from each nuclide to find the trt:1 air dose from beta radiation.

3.2.2 Method II The dose calculated shall follow the guidance of Regulatory Guide 1.109 using the meteorological dispersion parameters applicable during the periods of release.

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Table 3.2 I

Noble Gases and Dose Factor to Air 4

(to be used for Technical Specification 3.11.2.2)

Beta Air Dose Factor Gamma Air Dose Factor 3

3 1 (mrad-m )

. (mrad-m )

9p Nuclide DF pCi-yr pCi-yr Kr-83m 2.88E-04" 1.93E-05 Kr-85m 1.97E-03 1.23E-03 Kr-85 1.95E-03 1.72E-05 Kr-87 1.03E-02 6.17E-03 Kr-88 2.93E-03 1.52E-02 Kr-89 1.06E-02 1.73E Kr-90 7.83E-03 1.63E-02 Xe-131m 1.11E-03 1.56E-04 1

Xe-133m 1.48E-03 3.27E-04 Xe-133 1.05E-03 3.53E-04 Xe-135m 7.39E-04 3.36E-03 Xe-135 2.46E-03 1.925-03 Xe-137 1.27E-02 1.51E-03 Xe-138 4.75E-03 9.21E-03 Ar-41 3.28E-03

-9.30E-03

• 2.88E-04 = 2.88 x 10-4

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I 3.3 Technical Specification 3.11.2.3, Dose to an Individual This section is to be used to insure compliance with the following Technical Specification:

LIMITING CONDITION FOR OPERATION 3.11.2.3 The dose commitment to an individual from radioiodines, radioactive materials in particulate form and radionuclides with half-lives greater than 8 days other than noble gases in gaseous effluents released to unrestricted areas shall be limited to the following:

a.

During any calendar quarter 17.5 mrem, and

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b.

During any calendar year 115 mrem 3.3.1 Method I To insure that the dose limit to any organ is met, it is necessary to calculate the dose to the thyroid.

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3.3.1.1 Dose to the Thyroid Dthyroid (mrem) = 0.003 Q33 + 100 Q131I where:

Number of curies of Iodine-131 and Tritium released.

Q1311' Q3H =

This is to be based on present measurements.

3.3.1.2 Application of Method I Step 1.

Based on present measurements, determine the number of curies of Tritium, and Iodine-131 released during the period.

Step 2.

-Find the thyroid dose by performing the above multiplications.

I Step 3.

Add 2.9 x 10-2 (mrem / month) to account for doses received from Carbon-14 Step 4 Record the thyroid dose and maintain a cumulative record for the quarterly and annual periods.

3.3.2 Method II The dose calculated shall be in conformance with Regulatory Guide 1.109 using the meteorological parameters applicable during the periods of release.

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f 4.0 METEOROLOGY Meteorological data for the period May 1977 through April 1978 were analyzed for the values of and locations of the greatest offsite annual average dilution factors. The three maximum dilution factors occurred in one location at the restricted area boundary in the southwest sector, 762 meters from the primary vent stack. The dose models assume that all pathways occur at this location, which is a conservative assumption since actual residences, gardens and milk and meat animals do not occur here but at locations with lower dilution factors.

The dilution factors used are presented in Table 4.1; and the description of the model is presented in Appendix B.

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TABLE 4.1 i

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Meteorological Dilution Factors i

i Annual Average Non-depleted Dilution Factor (x /Q):

1.104 x 10-5 sec/m3 Effective Gamma Dilution Factor (fx /Qf )

3.2 x 10-6 sec/m3 I

Deposition Rate (D/Q)i 1.326 x 10-8

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5.0 ENVIRONMENTAL MONITORING The Radiological Environmental Monitoring stations are listed in Table 5.1.

The location of these stations with respect to the Yankee Rowe facility are-shown on topographic maps in Figures 5.1 and 5.2.

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IABLE 5.1 RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS

• EXPOSURE PATHWAY SAMPLE LOCATION DISTANCE FROM DIRECTION FROM

~AND/OR SAMPLE AND DESIGNATED CODE THE PLANT (KM)

THE PLANT

1. AIRBORNE Radiciodine and Particulate AP/CF-11 Observation Stand 0.5 NW AP/CF-12 Monroe Bridge 1.1 SW AP/CF-13 Ford Hill Road 2.7 SSE AP/CF-14 Harriman Power Station 3.2 N

AP/CF-21 Williamstown 22.2 W

2. DIRECT RADIATION GM-11 Observation Stand 0.5 NW GM-12 Monroe Bridge 1.1 SW GM-13 Ford Hill Road 2.7 SSE i

GM-14 Harriman Power Station 3.2 N

GM-15 Whitingham Line 3.5 NNE GM-16 Monroe Hill Barrier 1.8 S.

GM-17 Cross Road 3.5 E

GM-21 Willi amstown 22.2 W

3. WATERBORNE

a. Surface WR-ll Bear Swamp Lower Re s',

ir 6.3 Downriver WR-21 Har.

Reservoir 10.1 Upriver

b. Ground WG-ll' Plant I le on-site well WG-12 Sherman opring 0.2 NW

c. Sediment from Shoreline SE-11 d4 Station 36.2 Downriver SE-21 Harriman Reservoir 14.7 Upriver

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• Sample locations are shown on Figures 5.1 and 5.2

+ Station-lX's are indicator stations and Station-2X's are control stations. - _..

4 TABLE 5.1 (Continued)

RADIOLOGICAL ENVIRONMENTAL MONITORING STATIONS

• EXPOSURE PATHWAY SAMPLE LOCATION DISTANCE FROM DIRECTION FROM AND/OR SAMPLE AND DESIGNATED CODE THE PLANT (KM)

THE PLANT

4. INGESTION

a. Milk TM-ll Lively Farm 5.8 E

TM-12 King Farm 6.1 N

TM-13 Hicks Farm 8

SSE TM-21 Mt. Williams Dairy 21 WSW b.

Fish and Invertebrates FH-ll Sherman Pond 1.5 at discharge pond FH-21 Harriman Reservoir 10.1 Upriver c.

Food Products TF-ll Monroe Bridge 1.1 SW f

TF-12 Laffond Garden 2.9 S

TF-21 Willi amstown 21 WSW TV-11 Monroe Bridge **

1.1 SW 4

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• Sample locations are shown on Figures 5.1 and 5.2

• • TV-11 Station is for leafy vegetable.

+ Station-lX's are indicator stations and Station-2X's are control stations.

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APPENDIX A Basis for the Dose Calculation Methods A.

Liquid Release Dose Calculations There are three methods for calculating the doses resulting from liquid releases.

Method I provides a simplistic assessment of the doses resulting from normal operation of the Yankee Rowe Station while Method II provides a more accurate avsessment which approaches the accuracy of the Regulatory Guide 1.109 models. Method III is a complete evaluation of the resulting doses using the models of Regulatory Guide 1.109 with the parameters applicable during the time of release.

Methods I & II have 'een developed by applying the Regulatory Guide o

1.109 models to the Yankee Rowe Station for an assumed release of one curie of each nuclide released in -liquid effluents as reported in the Effluent and Waste Disposal Semi-Annual Report. This resulted in a site specific dose conversion factor providing the number of millirems per curie released for each nuclide.

Using these dose conversion factors, doses were calculated for each month of the thirty month period from January 1976 to June 1978, using only the principle dose contributing nuclides. These doses were then compared to the doses calculated by the models of Regulatory Guide 1.109 using all nuclides reported to verify the accuracy of the Model I and Model II approximations.

Yankee Rowe has in the past evaluated, monthly, tha releases of Carbon-14 For the thirty month period there was an average release of 9.3 x 10-3

(.

curiss per yacr released as liquid cffluants. Evaluation of the dose

(

resulting from this release shows an average dose of 1.43 x 10-3 mrem /yr to the total body and 7.13 x 10-3 mrem /yr to the critical age, organ (child bone). This is an insignificant dose when compared to the limiting doses and has thus been ignored.

In performing the site specific dose calculations Regulatory Guide 1.109 parameters were used with a plant discharge flow rate of 308 3

ft /sec and a mixing ratio of 0.54 The only dose pathway identified at Yankee Rowe is consumption of fish from the Deerfield River.

1.

Dose to the Total Body The total body dose is approximated in Method II by using the site specific dose conversion factors for the five major contributors to the total body dose:

(

Method II 4

Dtb (mreia) = 3.89 x 10 Q33 + 0.0151 Q89sr

+ 2.30 Q90Sr + 9 97 Q134cs + 5.89 Q137cs where the following site specific dose factors are defined in mrem per curie released:

3.89 x 10-6= Adult total body dose for Tritium.

Child total body dose for Strontium-89.

0.0151

=

I 2.30 Adult total body dose for Strontium-90.

=

Adult total body dose for Cesium-134

-9.97

=

Adult total body dose for Cesium-137.

5.89

=

' Method II has been applied to the releases occurring.during January,

1976 to June 1978 cnd the comparison to the Regulatory Guida 1.109

(

calculation is shown in Figure A.I.

The Method II equation has been further reduced by determining the maximum releases of Tritium and Strontium, calculating the resulting dose and including their contributions as a constant.

It was also found that the Cesium-134 could be correlated to the Cesium-137 release. These considerations lead to:

tb = 1.59 x 10-4 + 10 Q D

137Cs where:

1.59 x 10-4=

Dose resulting from maximum observed monthly releases of Tritium and Strontium.

Effective for Cesium-134 and Cesium-137.

10

=

C1

(

The comparison of this approximation to the Regulatory Guide 1.109 calculation is shown in Figure A.2.

Since 1.59 x 10-4 mrem /mo is an insignificant contribution to the total body dose, it has been ignored with the result:

Method I Dtb = 10 Q137Cs The comparison of the method one approximation to the Regulatory Guide l.109 calculation is shown in Figure A.3.

2.

Dose to the Critical Organ A similar approach was used to find the dose to the critical organ, t,

It was found that the following expression will provide a dose

(

that is close to the critical organ dose for each release in the period January 1976 to June 1978:

Method II 0.53 Q69sr + 9 4 090Sr Dorgan (mrem)

=

+ 3.9 x 10-6 g3H + l 1 Q131I

+ 12.4 Q134Cs + 9.4 Q137CS where the following site specific dose factors are defined in mrem per curie released:

0.53

= Child bone for Strontium-89.

9.4

= Adult bone for Strontium-90.

3.9x10-6

= Adult total body for Tritium.

f' 1.1

= Adult thyroid in Iodine-131.

12.4

= Teen liver in Cesium-134 9.4

= Teen liver in Cesium-137.

The child bone, adult thyroid or teen liver have all been observed to be the critical organ in some month of the thirty month period examined. Figure A.4 shows the comparison of the dose to the critical organ calculated by Method II and by the Regulatory Guide 1.109 models.

The dose to the critical organ was further simplified by combining the highest observed. releases of Tritium, Strontium-89 and Strontium-90 and calculating a monthly dose of 6.2 x 10-4 mrem from these three nuclides; then recognizing that the Cesium-137 b

~26-

release was always higher than the Cesium-134 release by a factor of 1.5 to 21.

These considerations led to:

organ (mrem /mo) = 6.2 x 104+Q131I + 20 Q137cs D

As before the 6.2 x 10-4 mrem /mo was considered to be insigiricant and was dropped leading to:

Method I Dorgan (mrem) =Q1317 + 20 Q137 Figure A.S shows the comparison of the critical organ dose calculated by method I and by Regulatory Guide 1.109 models.

B.

Dose Rate to an Individual 1.

Definitions and Constants

(

The basic form of the equations to determine the dose rates are from Regulatory Guide 1.109 and use the following definitions:

3 DFB

= Total body gamma dose factor for nuclide "i" mrem-m f

poi-yr 3

DFS

= Beta skin dose factor for nuclide "i" mrem-m i

pCi-yr DFf

= Combined site specific skin dose factor 3

DF

= Gamma air dose factor for nuclide "i" mrad-m T

g pCi-yr j

3 D[g

= Beta air dose factor for nuclide "i" mrad-m pCi-yr mrad D,T

= Gamma dose to air gp mrad D,0

= Beta dose to air gp D

= Dose to skin from beta and gamma mrem skin,. - _

b

= D se rate to total body due to noble gases mrem tb yr b

= Dose rate to skin due to noble gases mrem akin yr b

= Dose rate to thyroid due to iodine and mrem thyroid particulates yr sec X/Q

= Annual average undepleted dilution factor 3

m see

[X /Q3

= Effective average gamma dilution factor 3

m S

= Shielding factor, dimensionless y

1.11

= Average ratio of tissue to air energy absorption coefficient, dimensionless 6

1 x 10

= Number of picoeuries per microcurie pCi C1 2.5

= 1.11 SF [xfq) (3 cf,3) 1 x 106 (pC1/ C1) 6

= 1.11(.7)(3.2 x 10-6)(1 x yo )

pCi-see 3

I Ci-m 6

11.

= 1 x 10 X/Q 6

= (1 x 10 )(1.1 x 10-5) 4 3.17x10

= Number of picoeuries per curie divided pCi-sec by number of_ seconds per year Ci-yr 102.

= Site specific Iodine-131 dose conversion factor mrem for infant thyroid Ci

/

3 0.10

= 3.17 x 104 (pci yr ) [X Q) (sec/m )

Ci see

- (3.17 x 10 )(3.2 x 10-6) pCi-yr 4

3 Ci-m 0.35

= 3.17 x'104 (PCi L ) [x fq)

(3,cf,3)

Ci sec

= 3.17 x 104 (1.1 x 10-5) pCi-yr 3

Ci-m l.

2.

Dose Rate to the Total Body

(

The dose rate to the total body due to noble gases is calculated by methods derived from Regulatory Guide 1.109 as follows:

6 6tb (mrem /yr) = 1 x 10 S IX /Q3 E h WB p

i i

Sp = 0.7

[X /Q] = Annual Average Gamma Dilution Factor = 3.2 x 10-6 (sec/m )

3 6

= Release rate of noble gas "i" ( Ci/sec) 1 Otb (mrem /yr) = 2.24 E 6 ( Ci/sec) DFB 1

i i

3.

Dose Rate to the Skin The dose rate to the skin is calculated by methods derived from Regulatory Guide 1.109 as follows:

4 Dskin (mrem /yr) = 1.11 x SF x D,gp + 3.17 x 10 I

Qi x (X /Q) x DFSi where Qi = Release rate of noble gas 'i" (Ci/yr)

Now D,Iir = 3.17 x 104 [x /Q] I Q DF i

g Defining 6 in microcuries per see and combining gives 1

6 D[1 bskin (mrem /yr) = 1.11 x S x 1 x 106 [X /Q) I 1

p 6

+ 1 x 10 X /Q E ' h DFS i

1 i

where:

0( /QI = 3.2 x 10-6 3,cy,3

. X/Q

= 1.1 x 10-5 sec/m3 4

S

= 0.7 p

Substituting gives E

0 DFS 6 skin (arem/yr) = 2.5 I EFi + 11

'(

1 i

i,

k [2.5 DFg + 11 DFS 3

(

=

g g

Define j

DFf=2.5DF1 + 11 DFSg Then 6 DF[

bskin (mrem /yr) =E 1

i 4

Dose Rate due to Iodines and Particulates The dose rate due to iodines and particulates is based on limiting the dose rate to the infant thyroid due to Iodine-131.

ci 6 thyroid (mrem)=102(Ci-released)31.5(ci-sec)y131I(sec) mrem yr Ci-yr 3

ci hthyroid(mrem)=3.2x10(mrem-sec)6131I (see) yr Ci-yr

(

e

o C.

Dosa to Air due to Noble Gesas

(

The dose to air due to noble gases has been calculated by the methods of Regulatory Guide 1.109.

1.

Gamma Dose to Air The gamma dose is calculated as follows:

Y 3

D,gp(mrad) = 3.17 x 104 (PCi g) [ j (sec/m )

Ci see 3

E Q (C1) DF (mrad-m )

i i

i yr-pCi where:

[]

= Annual Average Gamma Dilution Factor = 3.2 x 10-6 (3,cf,3)

Q

= Number of curies of noble gas "i" released i

which leads to:

D,Y p(mrad) = 0.10 I Q (Ci) DF i

i 2.

Beta Dose to Air The beta dose to air is calculated in the same manner as the gamma dose except the undepleted X/Q is used (1.1 x 10-5 3

sec/m ) and the dose factor from beta radiation is applied, leading to:

8 Dair(mrad) = 0.35 I Q (Ci) DFi i

t D.

Dose to an Individuel Dun to R?diciodines end Perticulates

(

There are two methods for calculating the dose due to radiciodines and particulates. The first method was derived in a manner similar to that for the liquid pathways while the second method is calculation by the models of Regulatory Guide 1.109.

The pathways considered were:

inhalation, stored vegetables, leafy vegetables, goat milk, reat and direct exposure from the ground plane.

Regulatory Guide 1.109 parameters were used with a goat assumed to be on pasture 50% of the time consuming 80% of its feed from pasture during that period. It was found that the thyroid dose to the child or infant was controlling in every case.

An evaluation of thirty months of data on Carbon-14 releases through the gaseous pathway shows an average release of 1.53 x 10-2 Ci/mo which

(

would produce a maximum dose to a child's bone amounting to 2.87 x 10-2 mrem /mo.

1.

Dose to the Thyroid The dose to the thyroid is approximated in Method I by using the site specific dose conversion factors for the two major contributors to the thyroid dose:

Method I

+ 100 Q131I Dthyroid (mrem) = 0.003 Q3H where:-

.003 = ""'" for Tritium for child's thyroid.

( i

100 = mrem for Iodine-131 for infant's thyroid Ci Method I has been applied to the releases occurring during the period January 1976 to June 1978 and the comparison to the

{,

Regulatory Guide 1.109 calculation is shown in Figure A.6.

l i

f i '

s i

t t

6

.l -

^

n YANht.E R0WE LIQUID DOSES TO TOTAL BODY a= REG GUIDE 1 109 X: METHOD II 1E-02 E=

i

'l 1

C S

i G

E a

^

o IE i t

E

{

N

?

$2

,E Hh w tti T Of ao s>

E mL 3

s LLI

?

(n R

o 1E-04 3

g O

5 8

l!

C

. Cp i

.O O

lEllr i

i IE in g. p ng n p o piin.ip opoingn pi inig ni n p upiipiin ne n o no no"P"It"Im p o n nin.p n p onnin.p n nn oni i

1 9'16 1977 1978 C "J Z3>

r-

3SN > "

rh5 a8U

,_O $gr a*

I" "f"

"ig Y

D I

O B

4" l

L i"

A T

EO WT "i

I' 0

RO i

T E

s dS i

hE i

NS AO 0

ie YD

=

i o"

0 i

D 1

i" I

p 9+

U 0

Q 14 I

0 p

1 -

L E

i n

E9 p

D5 o

I.

p U1 G

i

=

C y

E0 n

\\

R b

up

=:

AX

.p

.p up E5 _ 'l-ll 1~_

Eii?l g

51i 45 E-1f i

2 3

4 5

0 0

0 0

n E

E E

E 1

I 1

1 orNrWMrw w(OOC toc:M lll _

3 3C l

OY

FIGURE A.3

- QUID DOSES TO TOTAL BODY, ?fETff0D I

(

~.=

=

a f'-

~@

b~

+

z

~

C E

m m

L C

F--

LLJ O

^

2F-E O

?

%Q 1

F-

=

LLJ

= cn LJ U) i~

i z to CO i

>- O

=

~

=

Q L

O

]

=

O

~_

O s

a m_

LLJ E

O *~

"3 O

O

. :. t~

OO O CD I

_~

ON

=.

LLI LL) m

=

II II

=

4X

=

--y_

l'lil8 U '

8"H 3 8 3 3 3 l8pp p,_. _.

____,,I"" ' 3 3 5 8 jr p p g e,

N M

W u3 O

O O

O l

I I

t

(

W W

W W

~

P00R 0RIBR\\L

twunu, noo

~36-

FIGURE A.4 LIQUID DOSES TO CRITICAL ORGAN, METHOD II l

)

=

L

=O

O Z

'/

=-

C E

O i

M O

.J 5-T 3.

O i

s-<

m-1 Z *-*

i CM

[

%O

-N

.c e io i

JW E~

Y Z U) i C LU E-pm e

Q O

T C

O i,,

6-*

~

D O

E n-e m m-1 Q-

,e OO Em I

OH Lt.1 LtJ 2-Q" E il II 4+

5 "H'

pppp ui.uspppp =,

nupppp, "s ti ti ftsp p.

e N

M T

O O

O O

O O

I I

I I

I k

W W

W W

d wew> nm P00R BRGINAL

(RNKuE R0WE LIQUID DOSES TO CRITICAL ORGAN A; REG GUIDE 1.109

+ atiE THOD I IE-01

?,

S I

1E-02 j g

s b

C o

s 95 e(W 9>

i 23 r

'r 1E-03 r-t

?

E

~

o" l

M w

n L1J O

5 O

i d

1E-04 4 5

O

?

T i

C">

C:3 ll2:3

$..g..g.........g........g.............p.........g.g.,g,g.,g..g..g.....g..i..,...,..............g..g..,,,,,...g.,g.,i 1E-05 1976 197/

1976 Z

3::=

r--

Fl C L"rt E A. 6

)

GASEOUS DOSES TO TilYROID, METHOD I 7

b

-:m

. h

m I:

h O

?

w O

E E

i E-LLJ Z 3W i

O E-MO H

c-

_~

,d

(

J Lij

m r y) c-E zO fl i

CO

~

E

~

V)

?

')

E-d UJ T

T

~

C3 H

l OZ JC

?

• --In e-.

IZ~

E O-O

m O' CD Q E-Q OOY

~ -. w

• Ltj 11 l' il 9

4GO E.

1 g a ii v i s s iung.n.puun.puun.i p ii.. < i s i g. p. i n s i i.u.i.in yu.a..puin. s p i s i s.. a i l-N m

O O

C l

i I

(

uJ W

W e

e e

wwme ma P00R ORGNAL

APPENDIX B 7

Basis for the Atmospheric Dilution Factors Annual average dilution factors based on onsite meteorological data were computed for routine (long-term) releases by the Yankee Atomic Electric Ccmpany's (YAEC)AEOLUS(2) Computer Code.

AE0LUS is based, in part, on the straight-line airflow model as discussed in Regulatory Guide 1.111(3) and includes the following basic features:

hourly meteorological data input (wind direction, wind speed, vertical temperature difference, and, optionally, direction fluctuation, air temperature, sea water surface temperature and solar radiation).

Straight-line air flow model with Gaussian diffusion, plume centerline and sector-average models with single or split (vertical / horizontal, i.e., split-sigma) atmospheric stabilities.

part-time ground level and part-time elevated releases (split-H model).

seabreeze option for coastal sites (split-H, split-sigma, trapping and fumigation).

multi-energy sector-averaged finite cloud dilution factors for gamma dose calculations (both normal and seabreeze cases).

terrain features.

plume rise.

depletion in transit (2 models),

recirculation correction factors (built-in options for flat terrains and river valleys, or user selected values).

deposition rates (2 models), and dose statistical distributions for postalated accidental radioactive releases and exposure intervals (based on dose rate data per unit dilution factor 0(/Q) as input; thyroid, total body beta, total body gamma and skin doses).

AE1LUS produces hourly and long-term averages of non-depleted dilution factors for evaluating ground level concentrations of noble gases, tritium, carbon 14 and non-elemental iodines, depleted dilution factor for estimating ground level concentrations 4

of elemental radioiodines and other particulates.

effective gamma dilution factors for evaluating gamma dose rates from a sector-averaged finite cloud (mu'.tiple-energy undepleted source),

and deposition factors for computing dry deposition of elemental radioiodines and other particulates.

A more detailed description of the AEOLUS diffusion model is provided I4) in section 2.3.5 (long-term diffusion estimates) of the NEP 1&2 PSAR I2) and the AEOLUS computer code manual Annual average non-depleted dilution-factors, effective gamma

(

dilution factors and deposition (D/Q) rates for Yankee Rowe were calculated

- _ ~ -.

l using the following AEOLUS options; Sector-average model with atmospheric l

stabilities based on temperature difference, split-H model, plume rise, and no recirculation correction factors. X /Q and D/Q values for the i

restricted area boundary critical sector are provided in Table 4.1.

The average dilution factors computed as described are subsequently classified, for each 22-1/2 degree sector, into groups, and corresponding cumulative probability distributions are prepared. For each sector the dilution factors at a number of percentile points are determined. These points define the percent of time a dilution value is equalled or exceeded.

4 l

6-I s

f i

1 i

t

- - _ -., ~

8 APPENDIX C i

Basis for Setpoint Determinations Gaseous Effluent Setpoints The primary vent stsek noble gas effluent monitor is an offline system consisting of a beta scintillator, electronics, analog ratemeter 6

readout and a strip chart recorder which' spans 10 to 1 x 10 cpm. The manufacturer (NMC) provides calibration data which indicates the response of the beta sensitive system for the various radionuclides likely to be The manufacturer's calibrations have been independently verified detected.

by direct calibration at the plant. Aged radwaste gas containing primarily Xe-133 and Kr-85 was introduced into the monitor in several concentrations and a response curve generated. System operation nas demonstrated a normal I

background of 20 cpm and 3 x 10-8 uci/cc of Xe-133 yields 1 cpm.

Quantitative analysis of noble gases in the routine effluent is performed to get the isotopic distribution for release calculations.

Policy 1.

All noble gas releases occur by the primary vent stack monitor pathway.

2.

The noble gas stack monitor and analog ratemeter will have an alarm setpoint which addresses the most conservative limit of Technical i

Specification 3.ll.2.1.a.

3 Technical Specification 3.11.2.1 does not define "at any time", but a time interval of.one hour shall be considered appropriate for plant procedures.

6tb = 2.24 6 I f DFB t

i i

i + 11 6 E T

bskin = 2.5 6 I f DFS f DF t

g g

b

= 500 mrem /yr gn bskin = 3000 mrem /yr 6

= Total noble gas release la Ci/sec 6/F

= Total noble gas release concentration at the point of effluence from the primary vent stack

= UCi/sec gross gas

= u Ci cc/sec stack flow cc f

= This fraction of noble gas "i" in the mixture I fi = 1.0 i

i rg

= The response of the noble gas detector to radionuclide "i" with respect to Xe-133. r133Xe

• 1 R

= The current calibration coefficient for the noble gas monitor in epm per uCi/cc of Xe-133 7

Typically R = 3.0 x 10 epm /p Ci/cc Xe-133 Step 1.

Using f values from laboratory quantitative analyses of noble g

gas source terms, compute 6 the gross noble gas release rate limit.

Step 2.

Compute the concentration in the vent stack corresponding to this release rate limit, 6/F, where F is the vent stack flow rate.

Step 3 Computer the noble gas stack monitor setpoint corresponding to theconcentration6/F.

Setpoint(COM) = Q x R g pi pi F

i Step 4 Instrument setpoint shall be 0.75 of calculated setpoint.

k.

w-

--n,-

my y

y

Liquid Efflutnt Setpoints i

The monitored liquid pathways at the plant are the liquid radwaste discharge (test tanks) and steam generator blowdown water.

Both waste streams discharge into the service water header which itself discharges into the condenser cooling water effluent.

The in-line radiation monitors are identical Nuclear Research Corporation "4W" samplers with 2"x2" NaI detectors.

Readout is on analog 0

ratemeters with a range of 10 - 10 CPM. Strip chart recorders give permanent records of the ratemeter indications.

Release Mechanics Steam generator blowdown water is collected in the blowdown tank and periodically released by automatic level controllers to the service water cooling loop in the auxiliary building. The tank fills at a nominal 5 gallons / minute and discharges at a nominal 110 gallons / minute. Thus, discharges of approximately 100 gallons each occur about 1 minute of each 20 during normal plant operaion. This pathway is considered a continuous release point.

Test tank liquids are released under the administrative control of a radioactive liquid discharge permit, prepared by the plant chemist.

The flow rate of the release is specified on the permit. Variable orifices allow releases up to 30 GPM. The release rate is adjustable and recorded on a strip chart during the release. Normally, liquid radwaste is discharged in 7000 gallon batches at 30 gallons per minute. Test tanks are filled, A test tank release takes analyzed and released about once every four days.

k,

I ebout 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; thus, this pathway is considered a b*tch release.

i Flow Rates The condenser cooling flow is assumed to be 138,000 GPM with two pumps operating and 69,000 GPM with one pump operating. Throttling of condenser cooling water is not practiced at Yankee Rowe.

During shutdown periods, the 4000 GPM service water provides dilution water flow. Flow rate is variable and estimated by pump curves.

The discharge rate from the steam generator blowdown tank is fixed by piping geometry and a relatively constant head on the tank. Estimation is done periodically by measuring the time for the tank level to decrease during a normal release, The discharde rate for the test tanks is controlled by the discharge f

line vari-orifices and limited to 30 GPM.

Radiation Monitors The radiation monitors were installed in November 1968. Initial isotopic calibrations of the detectors were performed at that time.

It was established that the gross gamma detector response was fairly independent of the gamma energy, as expected. Thus, the response is a function of the radioactivity concentration, and the gamma yield of the mixture, but not the gamma energies of the mixture. The electronics of each monitor channel has an adjustable alarm setpoint.

A control room alarm will sound and the effluent discharge valve will close whenever the following occurs.

ratemeter indication above the setpoint 1

loss of dstector high voltege loss of detector signal loss of power to channel i

instrument controls not set in OPERATE position Both release pathways operate sicultaneously. Therefore, the setpoint determination will address this.

Steam generator water is continuously composited and the composite analyzed weekly for gamma emitters. Each test tank batch release is analyzed for tritium, noble gases, and gamma emitters prior to release.

Nuclides which are detected above the LLD concentration are considered present for reporting and radiation monitor setpoint determinatin.

Normal Flow Paths Radwaste

, Steam Generator Test Tanks Blowdown fC22 fCl1 F x MFC i

"MPC" fraction for radiation monitors is -

t 1-CH C,3,3 l

g l

3E-3 2E-4 MPC for SG blowdown pathway is taken as 3E-7, the MPC for I-131.

l l

MPC for test tank pathway is the most restrictive.

l l

MPC for all gamma emitting isotopes, excluding noble gases, determined to i

. i

be pres 2nt in the test tank liquid in concentrations above the LLD of SE-7 pci/ml.

F dilution flow due to circulating and service water normal LG blowdown flow approximately 110 GPM.

f3:

normal test tank release flow 0 - 30 GPM.

f2=

C1=

activity concentration in SG blowdown water. Assumed all I-131, negligible tritium, dissolved noble gases.

the measured concentrations of isotopes in test tank batch releases C2=

of gamma emitting isotopes, excluding xenon-133. For determining flow restrictions on f, the actual concentration in uCi/ml is used.

2 For determining radiation monitor alarm points and estimated response during a release, the isotope concentration is multiplied by its gamma i

yield to give a " gamma" u C1/ml.

NOTE: Yankee Rowe has no degas capability on the liquid radwaste system.

The test tank liquid effluent monitor has the lower threshold discriminator set to reject pulses from gamma rays of 0.1 MeV and less. This allows the sizable contribution from xenon-133 to be eliminated. ' Otherwise, abnormally high responses would be obtained by this monitor due to Xe-133 alone.

Normal Operation with No Test Tank Release Test tank effluent trip valve and manual valves are closed. Test tank effluent monitor flushed with demineralized water and detector indicating normal background.

k -

.y.

~

-1

fC1 1 = F x MPC Then f2=0 and

(

x 3E-7 pCi/ml for I-131 C3 = F/f3 And the setpoint is S = C K counts per minute above background.

1 K=

calibration factor CPMb Ci/ml for the steam generator blowdown effluent monitor.

Operation with Both Effluent Paths Discharging Simultaneously Step 1.

Evaluate steam generator blowdown liquid activity. Use previous composite sample analysis results and steam air ejector radiation monitor to estimate blowdown water radioactivity, C.

1 i,

Step 2. ' Estimate release flow of test tank required, f

• 2 Step 3.

Verify pre-release calculations A

A A

+ gases + g y

blowdown)< 1

(*d3 3E-3 2E-4 MPC 3E-7 A=

activity of the identified radionuclides af ter dilution by flowrate F.

Step 4 Calculate the concentration of gamma emitters which would cause E A/MPC = 1 and use as the basic of the setpoint for the test tank effluent monitor...

9 T

m

TABLE C.1 Calculation of Setpoint March 1979 Yankee Rowe Data Total Body Skin Mixture Dose Factor Dose Factor Detector Fraction Times Times

Response

Mixture Fraction Mixture Fraction r

f i

i x ri fi Kr-85m 0.02 2.3E-05 3.0E-05 0

0 Kr-85 0.0003 SE-09 4E-07 1.15 40 l

Kr-87 0.01 5.9E-05 1.0E-04 1.5 0.015 Kr-88 0.03 4.4E-04 7.1E-05 1.15 0.045 Xe-131m 0.04 3.6E-06 1.9E-05 0

0 Xe-133=

0.01 2.5E-06 1.0E-05 0

0 Xe-133 0.48 1.4E-04 1.5E-04 1.0 0.48 Xe-135 0.18 3.2E-04 3.2E-04 1.3 0.22 l-Xe-138 0.01 8.8E-05 4.1E-05 1.5 0.011 0.80 Summation 1.0 E = 1.7E-03 E = 1.1E-03 i = 10 i

i 1

131,000 uCi x 3.0E7 cpm x 0.8 I"

uci Q = 500 mrem / year = 131,000 Serpoint =

see 1.0E7 c 2.24x1.7E-03 sec

, 3000 arem/ year

= 170,000 " '

314,000 GPM

=

see (2. 5x1.8E-03)+(11x1. 8E-03) 170,000 uci x 3.0E-7 C{j2 x 0.8 Setpoint =

1.0E7 408,000 CPM

=

(,_

Instrument Setpoint = 0.75 x 314,000

= 230,000 CPM.

+-

-a 9-4

,,4--,'te7-g-g g-

.--g-g iv.-

m-49e

(

APPENDIX D 1

Radwaste Equipment Sections 3.11.1.3 and 3.11.2.4 concern liquid and gaseous waste treatment; this appendix is provided to aid in understanding these systems.

The normal liquid effluent treatment system, the associated release pathways and the instrumentation for monitoring the liquid effluent releases are shown in Figure D.1.

Figures D.2 and D.3 are similar illustration for the gaseous effluents.

1

(

t l

l l

l l

t A

, l

^

f

^

I Primary Drate T

Collecting Tank T

compostte

"'- 30)

Hydrogenated Waste Streams Sampler Liquid Radwaste Steam Cenerater Treatment System L.P. Surse Tank er rerutcatt'en les Eschenger Outlet 31oudoun Weste

.i

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Ta nk c

t':ta toops and Presserteer Dratae

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Effluent Monitor w

I (TK-31)

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g g

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I (TK-32)

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?

Solte Weste f

(Drue Roller)

It i

Primary Boliding Turbine Building T

Swep Tank 2

Floor Drains g

(TE-24)

Test Tanks f.

Aerated Waste Streams (TK-34-1) l (TK-34-2) g' Leakase Free Charging Systen Pump Seele 8

Floor Drain Susp Free Lower PAB Composite Drains Free Component C.W. Surge Tank 6 Needere Sa ler Service 9143. Containment Sump Radioactive Lab Susp Cravity Drain Drains From Fuel Pit. Safet y Injeet ton 6 Shield

\\

C Tank C

Tank Cavity g

(

Effluent Monitor (TK-27)

Drains Free Ion Enchange Pit Drain Free VC Drata From Blowdown er Flash Tanks Drain From Primary Vent Stac k Floor Drains From PAS Drains From PAS Pipe Trench me L

Manitored Weste t

m.m.e

,.ek.

g Yankee Rowe (TK-29-1) n (n 2) g Normal Liquid Radwaste circulating Water Discharge Deertteld River Effluent Pathways (Sherman Pond)

(Figure - D.1) i

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Q

D

^

N L.P. Surge Tank Vent y

Main Coolant Vent Header

/h JL I

st 46 Activity Dilution Waste Holdup Frimary Drain Decay Tank Tank Collecting Tank TK-32 TK-31 TK-30 te drygen Analyser Compressor Suction Cooler Compressor Easte Cas surge i

E-19 i

K.O. Drum t

Drum r Ventilation System Header M

l

~

L r

O 3)

Compressor

(__

Discharge Q

Cooler Weste Gas Compressors'

~

C-3-1 C-3-2 b

"a3:=

F Water loop Seal FIG. D.3 YANKEE h0WE i

COVER GAS SYSTEM i

i

u b

(

References l

1.

Regulatory Guide 1.109, "Carculation of Annual Doses to Man From Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10CFR50, Appendix I", U.S. Nuclear Regulatory Commission, Revision 1, October 1977.

2.

Hamawi, J.N., "AEOLUS - A Computer Code for Determining Hourly and Long-Term Atmospheric Dispersion of Power-Plant Effluents and for Computi.ng Statistical Distributions of Dose Intensity from Accidental Releases", Yankee Atomic Electric Company Technical Report, YAEC-1120, January, 1977.

3 Regulatory Guide 1.111 " Methods for Estimating Atmospheric Transport

(

and Dispersion of Gaseous Effluents in Routine Releases from Light Water Cooled Reactors", U.S. Nuclear Regulatory Commission, March 1976.

4 NEP 1&2 Preliminary Safety Analysis Report, New England Power Company, Docket Nos. STN 50-568 and STN 50-569.

I l '-

.