ML17256A533

Rev 1 to Offsite Dose Calculating Manual

ML17256A533
Person / Time
Site:	Ginna
Issue date:	03/04/1983
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17256A532	List: ML17256A531 ML17256A533
References
PROC-830304, NUDOCS 8303090074
Download: ML17256A533 (62)

Text

Offsite Dose Calculatin Manual for Ginna Station Rochester Gas and Electric Corporation Revision 1

8303090074 830304 PDR ADOCK 05000244 P

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Ginna Station Offsite Dose Calculatin Manual TABLE OF CONTENTS Radiological Effluent Technical Specification Section Dose Calculation Manual

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kt 3.5.4 4.12.1 X.

Liquid Effluent Monitor Setpoints 3.5.4 4.12.2 II.

Gaseous Effluent Monitor Setpoints 3.9.1.1 XXI.

Liquid Effluent Release Concen-trations 3.9.1.2 IV.

Liquid Effluent Dose 10 3.9.1.3 3.9.2.3 5.5 V.

Liquid and Gaseous Radwaste Treatment 13 and Operability 3.9.2.1 VX.

Gaseous Effluent Dose Rate 17 3.9.2.2 VIX.

Gaseous Effluent Doses 18 4.10.1 VXII.'nvironmental Monitor Sample Locations 20 3.9.2.4 IX.

Preparation of Special Report to Demonstrate Compliance with Environmental Radiation Protection Standards 32 X.

References

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LIST OF TABLES AND FIGURES

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Table 1

Dose Parameters for Radioiodin'es and Radioactive Particulate, Gaseous Effluents 7

Table 2

Dose Factors for Noble Gases and Daughters Table 3

Dispersion Parameter (X/g) for Iong Term Releases, Plant Vent 25 Table 4

Dispersion Parameter (D/Q) for Long Term Releases, Plant Vent 26 Table 5

Dispersion Parameter (K/Q) for Long Term Releases, Containment Purge 27 Table 6

Dispersion Parameter (D/Q) for I,ong Term Releases, Containment Purge 28 Table 7

Dispersion Parameter (K/g) for Long Term Releases, Ground Vent 29 Table 8

Dispersion Parameter (D/Q) for Long Term Releases, Ground Vent 30 Table 9

Pathway Dose Factors Due to Radionuclides Other than

'Noble Gases 31 Figure 1

Ginna Station Liquid Waste Treatment System 15 Figure 2

Ginna Station Gaseous Waste Treatment and Ventilation Exhaust Systems 16 Figure 3

Location of Onsite Air Monitors and Post Accident TLD's 21 Figure 4

.Location of Farms for Milk Samples and Ontario Water-District Intake 22 Figure 5

Location of Offsite TLD's Figure 6

Location of Offsite Air Monitors 24

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II Li uid Effluent Monitor Set pints

'he Ginna Technical Specifications, Section 3.5.4, require alarm and/or trip

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setpoints fo'r radiation monitors on each liquid effluent line (reference 1).

Precautions, limitations and setpoints applicable to the operation of Ginna Station liquid effluent monitors are provided in plant procedures P-9 and RD-13.1.

" Setpoint values are calculated to assure that alarm and trip actions occur prior to exceeding the limits of 10 CFR 20 at the release point to the unrestricted area.

For added conservatism, liquid effluent release rates are administratively set so that only small fractions of the applicable 10 CFR 20 maximum permissible concentrations can be reached in. the discharge canal.

The calculated alarm and trip action setpoints for each radioactive liquid effluent line monitor and flow determination must satisfy the following equation:

Equation (1):

cf

<C F+f Where:

C =

the effluent concentration which implements the 10 CFR 20 limit for unrestricted

areas, in pCi/ml.

c =

the setpoint, in pCi/ml, of the radioactivity monitor measuring the radioactivity concentration in the discharge line prior to dilution and subsequent release.

f =

the flow as measured at the radiation monitor location, in volume per unit time, in the same units as F below.

F =

the dilution water flow as determined prior to the release point, in volume 'per unit time.

Liquid effluent batch releases from Ginna Station are discharged through a

liquid waste disposal monitor.

The liquid waste stream (f) is diluted (by F) in the plant discharge canal before it enters Lake Ontario.

The limiting batch release concentration (c) corresponding to the liquid waste monitor setpoint is calculated from the above expression.

Since the value of (f) is very small in comparison to (F), the expression becomes:

Where:

Equation (2):

c C.Ff the maximum permissible concentration of gross beta, gamma activity above background in the circulatipg water discharge at the unrestricted area boundary (1 x 10 pCi/ml).

the dilution flow assuming operation of only 1 circulating water pump (200,000 gpm).

the maximum waste effluent discharge rate through the designated pathway.

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The limiting release concentration (c) is then converted to a setpoint count rate by use of the monitor calibration factor determined per pr'ocedure RD-13.1.

The expression becomes:

Equation (2a):

Setpoint'(cpm)

=

, c( Ci/ml)

Cal. Factor (pCi/ml per cpm)

E~xam le (Li9nid Radwaste Monitor R-18):

If one assumes, for example, that the maximum pump effluent discharge rate (f) is 30 gpm, then the limiting batch release conce'ntration (c) would be determined as follows:

(

C / l) +

1 x 10

( Ci/ml)

~ 200 000

(

m) 30 (gpm) c

< 6.7 x 10 (pCi/ml)

The monitor R-18 alarm and trip setpoint (in cpm) is then determined utilizing the monitor calibration factor calculated in plant procedure RD-13.1.

Assuming a calibration factor of 9.5 x 10

~(Ci/ml) cpm and a limiting batch release concentration determined above>the alarm and trip setpoint for monitor R-18 would be:

6.7 x 10

( Ci/ml) 4

=7x10 cpm 9.5 x 10

~(Ci/ml) cpm The setpoint values for the Containment Fan Cooler Monitor (R-16), Spent Fuel Pit Heat Exchanger Service Water monitor (R-20),

Steam Generator Blowdown monitor (R-19), the Retention Tank monitor (R-21),

and the All Volatile Treatment waste discharge monitor (R-22) are calculated in a similar manner using Equation (2),

substituting appropriate values of (f) and the corresponding calibration factor.

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Gaseous Effluent Monitor Set pints The Ginna Technical Specifications (reference

1) require alarm and/or trip setpoints for specified radiation monitors on each noble gas effluent line.

Precautions, limitations and setpoints applicable to the operation of Ginna Station gaseous effluent monitors are provided in plant procedures P-9 and RD-13.1.

Setpoints are conservatively established for each ventilation noble gas monitor so that dose rates. in unrestricted areas corresponding to 10 CPR Part 20 limits will not be exceeded.

Setpoints shall be determined so that dose rates from releases of noble gases will comply with the Technical Specifi-cation requirements of 3.9.2.1.a(i),

which stipulate that the dose rate for noble gases shall be

< 500 mrem/yr to the total body and

< 3000 mrem/yr to the skin.

The calculated alarm and trip action setpoints for each radioactive gaseous effluent monitor must satisfy the following equation:

Equation (3):

o

< ~iv f kK where:

setpoint in cpm Qiv =

release rate limit by specific nuclide in pCi/sec discharge flow rate in cfm units conversion factor (cc/sec/cfm) calibration factor (pCi/cc/cpm)

The, general methodology for establishing plant ventilation monitor setpoints is based upon a vent concentration limit (in pCi/cc) derived from site specific meteorology and vent release characteristics'dditional radiation monitor alarm and/or trip setpoints are'alculated for radiation monitors measuring 'radioiodines, radioactive materials in particulate form and radionuclides other than noble gases.

Setpoints are determined to assure that dose rates from the release of these effluents shall comply with Technical Specification 3.9.2.l.a (ii), which requires that the dose rate for all radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days shall be

< 1500 mrem/yr to an organ.

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The release rate limit for noble gases shall be calculated by the following equation for total body dose:

Equation (4):

Q..(pCi/sec) 500 (mrem/ r) 3 3

K. (mrem/yr per pCi/m )

X/Q (sec/m

)

i and by the following equation for skin doses:

Equation (5):

Q. (pCi/sec) 3000 mrem/ r)

(L. + 1.1M.) (mrem/yr per pCi/m )

X/Q (sec/m

)

3 3

Where:

K. = The total body dose factor "due to gamma emissions for earth identified noble gas radionuclide, (in mrem/yr per pCi/m )

from Table 2.

L. = The skin dose factor due to beta emissions fog each identified i

noble gas radionuclide, (in mrem/yr per pCi/m

'from Table 2).

M. = The air dose factor due to gamma emissions fgr each identified 3.

noble gas radionuclide, in mrad/yr per pCi/m from Table 2

(unit conversion c'onstant of 1.1 mrem/mrad converts air dose to skin dose).

X/Q The highest calculat.ed annual a'verage dispersion parameter for estimating the dose to the critical3offsite receptor from vent.release point (v).

(in sec/m ).

The X/Q. is calculated by the method described, in Reg.

Guide l.ill (reference 6).

,Q.

= The release rate of radionuclide (i) from vent (v) which iv results in a dose rate of 500 mrem/yr to the whole body or 3000 mrem/yr to the skin of the critical receptor, (in pCi/sec).

Historically, xenon-133 is the principal noble gas released from all vents and is appropriate for use as the reference isotope for establishing monitor setpoints.

The whole body dose will be limiting, and the Xe-133 release rate limit is calculated by substituting the appropriate values in equation (4).

After the release rate limit for Xe-133 is determined for each vent, the corresponding vent concentration l'imits are calculated based on applicable vent flow rates.

Annually-derived monitor calibration factors (pCi/cc per cpm) convert limiting vent concentrations to count rate.

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Containment Plant Vent and Air E 'ector Noble Gas Monitors (Monitors R-12 R-14 and R-15)

Monitor R-12 measures noble gas activity in containm'ent when it is isolated, or in the containment vent during purge releases.

Noble gases being released via the plant vent are detected by R-14.

Monitor R-15, on the air ejector, normally indicates only background noble gas activity; however it serves as one of the first indicators of primary-to-secondary leakage.

Additional noble gas monitoring capability for the containment, plant and air ejector vents is provided by high-range effluent monitors R-12A, R-14A and R-15A, respectively.

Noble gas monitor setpoints are conservatively set in Procedure P-9 to correspond to fractions of the applicable 10 CFR 20 maximum permissible concentrations (MPCs) for unrestricted areas.

Fractions are small enough so as to assure the timely detection of any simultaneous discharges'rom multiple release points before the combined downwind site boundary concentration could exceed MPC.

Additional conservatism is provided by basing these

'setpoints upon instantaneous downwind concentrations.

Release rates during the remainder of a given year, combined with any infrequent releases at setpoint levels, are likely to result in only a very small fraction of the 10 CFR 20 annual limits.

E~xam le:

(Plant vent monitot R-14)

Using Xe-133 as the controlling isotope for the setpoint and assuming a measured activity at 1.63 E-6 pCi/cc, a ratemeter reading of 16 cpm above background and a vent flow of 6.45 E4 cfm.

Xe-133 efficiency = Activit Net:Ratemeter Reading Xe-133 efficiency = 1.63E-6

= 1.023-7 ECi/cc 16 cpm

'I =~f (Ki)(X/gv) 500 (2.94E2)(2.7E-6)

= 6.3E5 pCi/sec

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6. 3E5 ( Ci sec) 6.45E4 (cfm) 472 1.02-7 cfm

'pm c = 2.1E5 cpm (R-14 is set at 1/10 of this value, per Pr'ocedure P-9)

Continuous radioiodine and particulate monitoring on the air ejector is not required.

Calculations and plant measurements have indicated that the iodine and particulate source terms via the air ejector compared to other airborne release

pathways, are negligible.

(See references 5, 7, and 8).

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TABLE 1 DOSE PA1VMTERS FOR RADIOIODINES AND RADIOACTIVE PARTICULATE GASEOUS EFFIUENTS" Radio-nuclide P.

3.

Inhalation Pathway (mrem/ r er Ci/m )

P.

good 8 Ground Pathways (m

~ mrem/ r er Ci/sec)

P.

1.

Radio-Inhalation Pathway nuclide (mrem/ r er Ci/m )

P.i F~od 8 Ground Pathways (m

mrem/ r er Ci/s H-3 Cr-51 Mn-54 Fe-59 Co-58 I

Co-60

~ Zn-65 Rb-86 Sr-89 Sr-90 Y-91 Zr-95 Nb-95 Mo-99 RU-103 Ru-106 Ag-110m 6.5E+02 3.6E+02 2.5E+04 2.4E+04 1.1E+04 3.2E+04 6.3E+04 1.9E+05 4.0E+05 4.1E+07 7.0E+04 2.2E+04 1.3E+04 2.6E+02 1.6E+04 1.6E+05 3.3E+04 2.4E+03 1.1E+07 1.1E+09 7.0E+08 5.7E+08 4.6E+09 1.7E+10 1.6E+10 1.0E+10 9.5E+10 1.9E+09 3.5E+08 3.6E+08 3.3E+08 3.4E+10 4.4E+11 1.5E+10 Cd-115m Sn-126 Sb-125 Te-127m Te-129m Te-132 Cs-134 Cs-136 Cs-137 Ba-140 Ce-141 Ce-144 Np-239 I-131 I-133 Unidentified 7.0E+04 1.2E+06 1.5E+04 3.8K+04 3.2E+04 1.0E+03 7.0E+05 1.3E+05 6.1E+05 5.6E+04 2.2E+04 1.5E+05 2.5E+04 1.5E+07 3.6E+06 4.1E+07 4.8E+07 1.1E+09 1.1E+09 7.4K+10 1.3E+09 7.2E+07 5.3E+10 5.4E+09 4.7K+10 2.4E+08 8.7E+07 6.5E+08 2.5E+06 1.1E+12 9.6E+09 9.5E+10

"-The listed dose parameters are for radionuclides that may be detected in gaseous effluents.

These and additional dose parameters for isotopes not included in Table 1 may be calculated using the methodology described in NUREG-0133, Section 5'.1 (reference 2).

TABLE 2 DOSE FACTORS. FOR NOBLE GASES AND DAUGHTERS'"

Radionuclide Kr-83m Kr-85m Kr-85 Kr-87 Kr-88 I

Kr-89 oo.,'r-90 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138

.Ar-41 Total Body Dose Factor K.

(mrem/ r er Ci/m )

.7.56E-02: "

1.17E+03 1.61E+01 5.92E+03 1.47E+04 1.66E+04 1.56E+04 9.15E+01 2.51E+02 2.94E+02 3.12E+03 1.81E+03 1.42E+03 8.83E+03 8.84E+03 Skin Dose Factor 3

(mrem/ r er Ci/m )

1.46E+03

-1.34E+03 9.73E+03 2.37E+03 1.01E+04 7.29E+03 4.76E+02 9.94E+02-3.06E+02 7.11E+02 1.86E+03 1.22E+04 4.13E+03 2.69E+03 Gamma Air Dose Factor 51.

3 (mrad/ r er Ci/m )

1.93E+01 1.23E+03 1.72E+01 6.17E+03 1.52E+04 1.73E+04 1.63E+04 1.56E+02 3.27E+02 3.53E+02 3.36E+03 1.92E+03 1.51E+03 9.21E+03 9.30E+03 Beta Air Dose Factor N.

3 (mrad/ r er Ci/m )

2.88E+02 1.97E+03 1.95E+03 1.03E+04 2.93E+03 1.06E+04 7 '3E+03 1.11E+03 1.48E+03 1.05E+03 7.39E+02 2.46E+03 1.27K+04 4.75E+03 3.28E+03

>The listed dose factors are for radionuclides that may be detected in gaseous effluents.

These dose factors for noble gases and daughter nuclides are taken from Table B-1 of Regulatory Guide 1.109 (reference 3).

A semi-infinite cloud is assumed.

'-'"+.7.56E-02

= 7.56 x 10-2

Li uid Effluent Release Concentrations Liquid batch releases are controlled individually and each batch"release is authorized and based upon sample analysis and the existing dilution flow in the discharge canal.

Plant. procedures RD-7 and RD-8 establish the methods for sampling and analysis of each batch prior to release.

A release rate limit is calculated for each batch based upon analysis, dilution flow and all procedural conditions being met before.it is 'authorized for release.

The waste effluent stream, entering the discharge canal is continuously monitored and the release will be automatically terminated if,the pre-selected monitor setpoint is exceeded.

(See Sec'tion I.)

If gross beta analysis is performed for each batch release in lieu of gamma isotopic analysis, then a weekly composite for principal gamma emitters and I-131 is performed.

Additional monthly and quarterly composite analyses are to be performed as specified in Table 4.12-1 of the Ginna Technical Specifications.

The equations used to calculate activity are:

Gamma S ectrometr eak area counts-bk d counts pCi/cc Act. =

(Count Time) (Eff. ) (Vol.) (T1/2 correction) (3. 7E4) (Branching fraction)

Gross Beta Gamma:

total counts - bk d counts (Count Time)(Eff.)(Vol.)(T1/2 correction)(3..7E4) where:

count time is in seconds; eff. = counting efficiency, in counts er sec disintegrations per sec.

vol. = volume, in milliliters; T1/2 correction

= decay correction factor, dimensionles s 3.7E4 = conversion constant, in disinte rations er sec; pCi Branching fraction is the fraction disintegrating by a particular decay mode, dimensionless.

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IV.

Ii uid Effluent Dose The dose contribution received by the maximally exposed individual from the ingestion of Iake Ontario fish and drinking water is determined using the following methodology.

These calculations will assume a near field dilution factor of 1.0 in evaluating the fish pathway dose, and a dilution factor of 20 between the plant discharge and the Ontario Water District, drinking water intake located 1.1 miles away (Figure 4).

The dilution factor of 20 was derived from drift and dispersion studies documented in reference 4.

Dose contributions 'from shoreline recreation, boating and swimming have been shown to be negligible in the Appendix I dose analysis (reference

5) and do not need to be routinely evaluated.

Also, there is no known human consumption of shellfish from Iake Ontario.

The dose contribution to an individual will be determined to ensure that it complies with the Technical Specification requirements of 3.9.1.2.a (i) and 3.9.1.2.a (ii).

The dose or dose commitment to an individual from radioactive materials in liquid effluents released to unrestricted areas shall be limited:

(i) During any calendar quarter to

< 1.5 mrem to the total body and to < 5 mrem to any organ, and, (ii) During any calendar year to

< 3 mrem to the total body and to < 10 mrem to any organ.

Offsite receptor doses will be determined for the limiting age group and organ, unless census data show that actual offsite individuals are of a less limiting age group.

The following expression is used to calculate ingestion pathway dose contributions m

for the total release period

> X l At from all radionuclides identified in liquid effluents released to unrestricted areas:

m Equation (6):

D

= X [A.

Z bt> C.> F<]

i,--, g =l Where:

the cumulative dose commitment to the total body or any organ, t, from the liquid effluents for the total time period (in mrem).

m b,t~,

'R= l C.iQ the length of the 2th time period over which C.< and if F> are averaged for all liquid releases, (in hours).

the average 'concentration of radionuclide, i, in undiluted

'iquid efflu'ent during time period ht< from any liquid

release, (in pCi/ml).

A.it the site-related ingestion dose commitment factor to the total body or any organ t for each identified principal gamma and beta emitter (in mrem/hr per pCi/ml).

See equation (7).

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F

=

the discharge canal dilution factor for C.

during any liquid effluent release.

Defined as the ratio of the maximum undiluted liquid waste fl'ow during release to the average flow from the site discharge structure to unrestricted receiving waters.

The dilution factor will depend on the number of circulation pumps operating and, duringI W3.Iitex'onditions, the percentage opening of the recirculating gate.

Reference curves are presented in plant procedure RD-7.

Equation (7):

A.

= k (U /D

+ U BF.) DF.

Where:

A.i'E, k0 the site-related ingestion dose commitment factor to the total body or to any organ t for each identified principal gamma and beta emitter, (in mrem/hr per pCi/ml).

units conversion factor, 1.14 x 10

= 10 pCi/pCi x 10 ml/kg 5

6..

3 8760 hr/yr.

U

=

a receptor person's water consumption by age group from w

table 'E-5 of Regulatory Guide 1.109 (reference 3).

D

=

Di.lution factor from the near field area of the release w

point to potable water intake.

The site specific dilution factor is 20.

This factor is assumed to be 1.0 for the fish ingestion pathway.

U

=

a receptor person's fish consumption by age group from table F

E-5 of Regulatory Guide 1.109.

BF. =

Bioaccumulation factor for nuclide",', (in fish pCi/kg per PCi/2), from Table A-1 of Regulatory Guide 1.109,.

DF. =

Dose conversion factor for the ingestion of nuclide, i, for a receptor person in pre-selected organ, t, (in mrem/pCi),

from Tables E-ll,'-12, E-'3, E-14 of Regulatory Guide 1.109.

The monthly dose contribution from releases for which radionuclide concentrations are determined by periodic composite sample analysis may be approximated by assuming an average monthly concentration based on 'the previous monthly or quarterly composite analyses.

However, in the radioactive effluent release report (submitted within 60 days of January 1 per Technical Specification 6.9.14) the calculated dose contributions from these radionuclides shall be based on the actual composite analyses.

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Exam le which illustrates how to com ute the dose to the whole bod via the fish and drinkin water athwa s

assumin an initial Cs-137 dischar e

concentration of 3.0 E-4 Ci/ml:

Given the following discharge factors, where:

bt~

Fu 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 3.0 E-4 pCi/ml 200,000 gpm D

=

.20 And, taking the following values from Regulatory Guide 1.109 which concern the receptor of interest, which we assume is the child in this case:

U

=

510 R/year w

U

=

6.9 Kg/year BF. =

2000 pCi/kg per pCi/E I.

DF. =

4.62 E-5 mrem/pCi i

Then, the site-related ingestion dose commitment factor, A.

, is calculated as follows:

= 1.14 E5

(

+ 6.9 2000) 4.62 E-5 4

510 20

.A.

=

7.28 E4 mrem/hr per pCi/mi ix And, the whole body dose to the child is then:

D mrem = (A. )(bt>)(C.>)(F>)

= (7.28 E4)(1)(3.0 E-4)(1.0 E-4)

D

= 2. E-3 mrem to the whole body from Cs-137 (The dose contribution from any other isotopes would then need to be calculated and summed).

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Li uid and Gaseous Radwaste Treatment and 0 erabilit I

An objective of the Ginna Technical Specifications which implement the overall requirements of 10 CFR Part 50, Appendix I, is to ensure that the plant radwaste treatment equipment is used and maintained.

This equipment is to be utilized to reduce radioactive discharges from nuclear plants to levels "as low as reasonably achievable" or ALARA.

AIARA levels warranting equipment operability have been defined by.the NRC in the form of monthly dose "trigger" values.

The trigger values, which are provided below, correspond to approximately 1/48 of the annual design objective doses given by 10 CFR Part 50, Appendix I. If continued at this rate, these monthly doses would correspond to just under 1/4 of the Appendix I annual design objectives.

Li uid Radwaste S stem Gaseous Radwaste S stem Ventilation Exhaust 31-day Trigger Values 0.06 mrem (w. body) 0.2 mrem (any organ) 0.2 mrad (gamma air) 0.4 mrad (beta air) 0.3 mrem (any organ)

Figures 1 and 2 show the Ginna liquid and gaseous waste/ventilation exhaust systems.

These systems are normally in routine use at the plant.

Because discharges are being treated, the trigger values in the Technical Specifications may be exceeded but compliance with the stated quarterly and annual dose limits is required.

If the liquid or gaseous radwaste/ventilation exhaust systems were to be inoperable'n excess of 31 days, then effluents are considered "untreated" waste.

Should, over a 31-day period, the plant discharges exceed the dose

system, then Technical Specifications 3.9.1.3.b and 3.9.2.3.c apply.

In this

case, a 30-day report must be submitted to the Commission which identifies the inoperable equipment and describes appropriate corrective actions (see Technical Specifications 3.9.1.3.b and 3.9.2.3.c).

~Exam le:

Assume a case where plant modifications were underway on the waste evaporator and demineralizer piping, and a reduction of the liquid waste volume contained in the waste holdup tank (WHUT) was needed.

Assuming no other means of treatment were readily available, WHUT liquid (untreated) might be transferred to the waste condensate

tanks, sampled and then discharged on a controlled basis.

In this case, assume also that a decision was made to proceed this way at the beginning of a given month, knowing that the waste evaporator and other portions of the liquid radwaste treatment system would be unavailable for the next 45 days.

The followin method would be used to determine the need for a 30-da re ort:

1.

Using existing plant procedures, sample the concentration contained in the WHUT (C.<).

Decide a sample frequency (e.g.

1-day) since the tank

.iQ concentration could change.

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2.

Determine the permissible release rate to maintain, the concentration in the discharge canal well within the applicable maximum permissible concentration (e.g.

1/100-1/10 MPC) for the mixture.

(The discharge canal concentration is equal to C.

~ F ).

Calculate the incremental dose from all identified isotopes via the drinking water and fish ingestion pathways for each receptor group.

The critical receptor will probably be the child.

Assume the release will be continuous and that doses will be evaluated each day, corresponding to the WHUT sampling frequency.

(We thus compute D

using Equations 6 and 7, taking bt as the duration of each release; in this case, 24 hr/day).

4.

The offsite receptor dose due to a controlled discharge of the WHUT contents is thus determined and cumulated over each daily release time interval. If the WHUT isotopic mixture and the discharge canal dilution fa'ctor, F>, are relatively constant, then, each day!s dose increment should be approximately the same.

One can then estimate the number of release days it will take to reach the applicable dose trigger value.

5.

The 30-day reporting requirement applies if a radwaste treatment system is inoperable and dose trigger values are exceeded.

In the example, one reporting criterion is already met, since the treatment systems will be out of service for more than 31'ays.

If we determine that the liquid pathway dose does not exceed the trigger values in 31 days or less, then a 30-day report is not required.

However, if a liquid pathway dose attains a trigger level within 31 days, then a xeport submittal would be required.

6.

In the last case, it would be prudent to avoid a situation requiring the 30-day report.

First, a trigger level, dose, when added to the calculated doses resulting from all other liquid release sources (e.g. high conductivity waste tank, blowdown, retention tank),

may significantly impact upon the plant's "dose budget" for the calendar quarter or the calendar year.

Also, more realistically, other treatment options would likely be available at the plant and could be utilized.

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GINNA STATION LIQUID WASTE TREATMENT SYSTEM FIGURE 1 BLOWDOWN SAMPLE LINE i

OMONXTOR R-19

'RECYCLE S/G BLOWDOWN LINE TO HOTWELL SPENT RESIN STORAGE TANKS REACTOR COOLANT DRAIN TANK CONTAINMENT SUMP "A" CHEMICAL DRAIN TANK LAUNDRY 6 HOT SHOWER TANKS I

S/G BLOWDOWN TANK DRAIN

• CI-AUX.

S(

INTERMEDIATE BLDG. DRAINS MIXED BED WASTE HOLDUP TANK WASTE EVAPORATO MIXED BED DI HIGH CONDUCTIVITY WASTE TANK MONITOR R-22 CIRCULATING DISCHARGE WATER,'ASTE CONDENSATE TANK TURBINE BLDG DRAINS SERVICE WATER MONITOR R-21 SFP HX CV FAN COOLER MONITOR R-20 MONITOR R-16 MONITOR R-18 WASTE

,CONDENSATE TANK

• OR TO CIRC.

WATER DISCHARGE

GINNA STATION

GASEOUS WASTE TREATMENT AND VENTILATION EXHAUST SYSTEMS FIGURE 2 AUXILIARYBUILDING;"

"G"FILTERS C.'ONITORS R-lOB,13,14,14Al VENTILATION SYSTEM "C"FILTERS F

A F

PLANT VENT

."A"FILTERS

,GASEOUS WASTE CVCS TREATPIENT. SYSTEM'ASTE i

GAS COMPRESSORS GAS DECAY TANKS 04 TO PLANT VENT MONITORS CONTA1NMENT PURGE"(

CONTAINMEN~T

'., C F

MONITORS R-lOA,11,12,12A CONTAINMENT VENT CONDENSER AIR EJECTOR;

~MONITORS R~5 R~5A j OFFGAS VENT NOTE: M=..HEPAMZLTERS C=CHARCOAL FILTERS F=FANS

II'1

)<

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t

VI.

Gaseous Effluent Dose Rate Gaseous effluent monitor setpoint's as described in Section II of this manual are established at concentrations which permit some margin for corrective action to be taken before exceeding offsite dose rates corresponding to 10 CFR Part 20 limitations.

Plant procedures RD-1.1, RD-1.2, RD-1.3, RD-2, RD-3, RD-5 and RD-12 establish the methods for sampling and analysis for continuous ventilation releases and for containment purge releases.

Plant procedure RD-6 establishes the methods for sampling and analysis prior to gas decay tank releases.

The dose rate in unrestricted areas due to unplanned releases of airborne radioactive materials may be averaged over a 24-hour period ac'cording to Technical Specification 3.9.2.1.b.

Dose rate shall be determined using the following expressions:

For noble gases:

Equation (8):

D = Z [K.

(X/Q)

Q. ]

< 500 mrem/yr (to total body) i Equation (9):

D = X [(L. + 1.1 M.) (X/Q)

Q.

] < 3000 mrem/yr (total gamma 6

1 v iv the skin)

For radioiodines, radioactive materials in particulate form, and radionuclides other than noble gases.

Equation (10):

D = Z [P.

W Q.

] < 1500 mrem/yr (critical organ)

K.i The total body dose factor due to gamma emissions for each jdentified noble gas radionuclide, (in mrem/yr per pCi/m from Table 2).

The skin dose factor due to beta emissions fog each identified noble gas radionuclide, (in mrem/yr per pCi/m from Table 2).

M.

The air dose factor due to gamma emissions fog each identified noble gas radionuclide, (in mrad/yr per pCi/m from Table 2),

(unit conversion constant of 1.1 mrem/mrad converts air dose to skin dose).

P.

The dose parameter for radionuclides other than noble gases for the inhalation/pathway, in modem/yr per pCi/m and for food and ground plane pathways, (in m

~ mrem/yr per pCi/sec) from Table 1.

The dose factors are based on the critical individual organ and most restrictive age group.

(X/Q)

The highest calculated annual average relative concentration for any area at or beyond the unrestricted area boundary, (in sec/m ).

Wv The highest annual average dispersion parame)er for estimating the dose to the critical receptor; (in sec/m for the inhalation pathway, and in m for the food and ground pathways).

Qiv the release rate of radionuclide i from vent (v), (in pCi/sec).

- 17 "

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VII. Gaseous Effluent Doses The air dose in unrestricted areas due -to noble gases released in gaseous effluents from the site shall be determined using the following expressions:

During any calendar year, for gamma radiation:

Equation (ll):

D

= 3.17 x 10 X[M. (X/Q)

Q. ] < 10 mrad, and

-8 i

v iv i

During any calendar year for beta radiation:

Equation (12):

D

= 3.17 z 10 Z [N. (X/Q)

Q. ] < 20 mrad 3.

Where:

M. = The air dose factor due to gamma emissions fog each identified

.i noble gas radionuclide, (in mrad/yr per pCi/m from Table 2).

N. = The air dose factor due to beta emissions for3each identified noble gas radionuclide, (in mrad/yr per pCi/m from Table 2).

(X/Q)

= For vent releases.

The highest calculated annual average v

relative concentration for any area at or beyond the unrestricted area boundary, including uninhabited areas, (in sec/m ).

D

= The total gamma air dose from gaseous effluents, (in mrad).

D

= The total beta air dose from gaseous effluents, (in mrad).

Q.

= The release of noble gas radionuclides, i, in gaseous iv effluents from all vents, in pCi.

Releases shall be cumulative over the time period.

-8 3.17 x 10

= The inverse of the number of seconds in a year.

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The dose to an individual from radioiodines and radioactive, materials in'articulate form with half-livesgreater than 8 days in ga'seous efflu'ents released from the site to unrestricted are'as shall be determined using the following expression:

During any calendar year:

Equation (13):

D> = 3.17 x 10 Z R.

[W

(}. ], < 15 mrad

-8 Where:

~iv = The release of radioiodines, and radioactive materials in particulate form in gaseous effluents, i, with half-lives greater than 8 days, (in pCi).

Releases shall be cumulative over the desired time period as appropriate.

D

= The total dose from radioiodines and radioactive I

materials in particulate form with half-lives greater than 8 days in gaseous effluents, (in mrem).

The annual average dispersion parameter for estimating the d~se to an individual at the critical logation; (in sec/m for the inhalation pathway, and in m for the food and ground pathways).

R. = The dgse factor for each identified radionuclide,3i, l.

(in m

~ mrem/yr per pCi/sec or mrem/yr per pCi/m from Table 9).

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.'ETERS

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CIRCULATING WATER DXSCHA'RGE 2.

FIGURE 4 LAKE ONTARIO

'RUSSELL STATION TIIAEE Mll.E roun MILE IIIIIIII TWO ONE MIl.E L nKE.

PINNA SITE T(

AOAP Ontario Water District'-

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TABLE 3 DISPERSION PARAMETER

(~X Q)

FOR LONG TERN RELEASES

> 500 HR/YR OR

> 125 HR/QTR Plant Vent Distance to the control location, in miles Sector 0-0. 5

0. 5-1. 0 1,. 0.>>2. 5 l. 5..2..0.

2.. 0.-2. 5.

.2..5-3..0 3.. 0.-3. 5

3. 5-4. 0

4. 0-4. 5 4.5-5.0 ENE SE SSE SSN T. SN Til

~ I 8.8 E-6 2.1 E-6 7.4 E-6 1.7 E-6 9.7 E-6 2.3 E-6 9.2 E-6 2.2 E-6 1.1 E-5 2.7 E-6 8.5 E-6 2.1 E-6 6.5 E-6 1.4 E-6 3.6 E-6 1.1 E-6 2.1 E-6 8.8 E-7 2.0 E<<6 5.8 E-7 2.3 E-6 5.6 E-7 2.9 E-6 7.1 E-7 3.3 E-6 1.0 E-6 2.7 E>>6 8.9 E-7 2.0 E-6 6.4 E-7 4.3.E-6 1.2 E-6 1.0 E-6 4.7 E-7 2.5 E-7 9.2 E-7 4,5 E-7 2.5 E-7 1.2 E-6 5.9 E-7

~ 3.2 E-7 1.1 E-6 1.3 E-6 5.0 5.4 E-7 2.6 E-7 E-7 2.7 E-7 4.5 E-.7 1.9 E-7 3.4 E-7 1.8 E-7 1.0 E-7 9.6 E-8 3.0 E-7 1.4 E-7 7.6 E-8 5.3 E-7 1.6 E-7 9.0 E-8 5.1 E-7 2.4 E-7 4.7 E-7 2.3 E-7 1.3 E-7 1.2 E-7 3.6 E-7 1.8 E-7 9.8 E-8 5.7 E-7 2.7 E-7 1.4 E-.7 1.1 E-6 4.4 E-7 2.2-E-7 6.9 E-7 -3.0 E-7 1;5 E-7 5.0 E-7 2

3 E-7 1.2 E-7 1.8 E-7 1.8 E-7 2.3,E-7 1.8 E-7 1.9 E-7 1.5 E-7 1.1 E-7 8.4 E-8 7.6 E-8 6.8 E-8 5.4 E-8 6.4 E-8 9.6 E-8 9.0 E-8

~

7.4 E-8 1.0 E-7 1.3 E-7 1.1 E-7 9.4 E-8 1.4 E-7 1.2 E-7 9.9 E-8 1;8 E-7 1.5 E-7 1.2 E-7 1.4 E-7 1.2 E-7 9.8 E-8 1.4 E-7 1.2 E-7 9.6 E-8 1.1 E-7 9.4'E-8 7.9 E-8 8.5 E-8 6.3 E-8

6. 9 E-8

5. 6 E-8 5:2 E-8 4.2 E-8 6.9 E-8 5.7 E-8 5.S E-S 4.S E-8 4.6 E-8 3.9 E-8 8.0 E-8 6.7 E-8 5.6 E-8 5:9 E-8 4.8 E-8 4.0 E-8 5.3 E-8 4.5 E-8 3.8 E-8 4.2 E-8 3.5 E-8 2.9 E-8 4.8 E-8 3.9 E-8 3.3 E-8 7.2 E-8 5.9 E-8 4.9 E-8 8.2 E-s.

9.0 E-8 1.1 E-7.

8.7 E 8.5 E-8 6.9 E-8 4.8 E-8 3.5 E-8 3.3 E-8 3+2 E

8

2. 4 E-8 2.9 E-8 4.3 E-8 4.9 E-s

• Direction wind blows into

TABLE 4 DXSPERSXON PARAMETER

{D Q)

FOR LONG TERM RELEASES

> 500 HR YR OR

> 125 HR QTR Plant Vent Distance to the control location, in miles Sector*

0-0.5 0.5-1.0

1. O.-l, S 1, 5..2..0.

2.. 0-2.. 5.

2. 5-3.,0 3,. 0-3.. 5 3. 5-4. 0 4. 0-4. 5 4. 5>> 5. 0 A

NE SSZ SSH SN ASH hR'7

.Pi%

MAW 8.3 E-8 4.5 E-8 6.5 E-8 8.3 E-8 1.4 E-7 1.4 E-7 1.3 E-7 5.8 E-8 2.8 E-8 3.1 E-8 4.5 E-8 5.6 E-8 4.2 E-8 2.2 E-8 1.5 E-8 4.0 E-8 1.7 E-8 1.0 E-8 1.5 E-8 1.8 E-8 2.9 E-8 3.0 E-8 2.7 E-8 1.4 E-8 8.6 E-9 7.8 E-9 1.0 E-8 1.3 E-8 1.0 E-8'.9 E-9 4.1 E-9 9.2 Z-9 6.1 E-9 2.S E-.9 3.7 E-9 1.S 5.4 E-9 2.2 E-9 6.4 E-9 2.6 E-9 1.0 E-8 4.2 E-9 1.1 E-8 4.3 E-9 9.3 E-9" 3.7 E-9 4.7 E-9 1.9 E-9 F 1 E-9 1.3 E-9 3.1 E-9 1.3 E-9 3.6 E-9 1.5 E-9 4.6 E-9 1.8 E-9 3.9'E-9 1.6 E-9 2.4 E-9 1.0 E-9 1.7 E-9 7.0 E-10 3.5 Z-9 1.4 E-9 1.2 E-.9 7.3 E-.10 7.0 E-.10 4.4 g-10

l. 0 E-9 6, 5:E-10

l. 2 E-.9

7. 5 E-10 1.9 E-.9 1.2 E-9 1.9 E-9 1.2 E-9 1.7 E-9 3..0 E-9 8.9 E-10 S.6 E-10 5.8 E-10 3;8 E-10 5.9 E-10 3.7 E-10 6.8 E-10 4.4 E-10 8.4 E-10 5.3 E-10 7.4 E-lo 4.7 E-10 4.7 E-10 3.0 E-10 3.3 E-10 2.1 E-10 6.6 E 10 4.2 E-10 5.1 E-.lo 3,1 E 10 4.5 E-.lo

5. 3 E-10

8. 6 Z-10 8.7 E-10 7;7 E-.10 4.1 E-'10

2. 9 E-10 2.7 E-10 3.1 E-10 3.7 E-10 3.3 E-10 2.1 E-10 1.5 Z-10 2.9 Z-10 4.1 E-10 2.9 E-.10 2.5 E-10 2.4 E-. 10 1.8 E-10 1.5 E-10 3.6 E-10 2.6 E-10 4.1 E-10 3.1 E-10 2.2 E-10 2.6 EA 1.7 E-10 1.3 E-3.0 1.2 E-10 8.8 E-ll 7.4 E%

2.3 E-10 1.7 Z-10 1.4 E-10 6.7 E-10 4.8 E-10 4.1 E-10 6.7 E-10 5.2 E-10 4.5 E-10 6.1 E<<10 4.6 E-10 4.0 E-10 3.5 E-10 2.7 E-10 2.3 E-10 2.4 E-10 1.8 E-10 1.6 E-10 2.2 E-10 1.8 E-10 1.5 E-10 2.5 E-10 1.9 E-10 1.6 E-10 2.9 E-10 2.1 E-10 1.8 E-10 2.6 E-10 1.9 E-10 1.6 E-10

TABLE 5 DISPERSION PARAMETER (X/Q>

FOR LONG TERM RELEASES

> 500 HR/YR OR > 125 HR/QTR c

'on'tainment Purge Distance to the control location, in miles Sector" 0-0. 5

0. 5-1. 0 l.. 0.-3., 5

1. 5.-2..0.'.. 0-2. 5.

2. 5-.3..0 3.. 0.-3.. 5

3. 5-4. 0

4. 0-4. 5

4. 5-5. 0 LXNw h'lE ES-SS" SSPi 3.7.E-6 1.2 E-6 3.1 E-6 1.0 E-6 4.1 E-6 1.4 E-6 3.9 E-6.

1.3 E-6 4.9 E-6 1.6 E-6 4.3 E-6 1.5 E-6 4.2 E-6 1.2 E-6 2.3 E-6 9.7 E-7 1.3 E-6 7.7 E-7 1.2 E-6 4.5 E-7 1.3 E-6 4.1 E-7 1.7 E-6 5.3 E-7 1.7 E-6 7.2 E-7 1.2 E-6 6.0 E-7 8.5 E-7 4.4 E-7 1.8 E-6 7.0 E-7 7.2 E-7 6.6 E-7 9.0 E-7 7.7 E-7 8.8 E-7 9.1 E-7 6.1 E 4.6 E-7 4.1 E-7 3.3 E-7 2.7 E-7 3.2 E-7 4.4 E-7 3.9 E-7 3.0 E-7 4.4 E-7

3. 6 E-7

2. 0 E-7 l. 4 E-7 3.5 E-7 2.0 E-7 1.5 E-7 4.7 E-7 2.7 E-7 2.0 '-7 3.9 E-7 2.1 E-7 1.5 E-7 4.1 E-7 2.2 E-7 1.5. E-7 3.9 E-7 2.0 E-7

-2.8 E-7 1.4 E-7 1.4 9.9 E-8 2.2 E-7 1.2 E-7 8.1 E-8 1.9 E-7 1.0 E-7 7.4 E-8 1.7 E-7 9.5 E-8 6.7 E-8 1.3 E-7 7.3.E-8 5.'2 E-8 1.5 E-7 8.6 E-8 6.0 E-8 2.1 E-7 1.2 E-7 8.6 E-8 2.0 E-7 1.1 E-7 8.2 E-8'.6 E-7 8.9 E-8 6.5 E-8 2.2 E-7 1.2 E-7 9.0 E-8 1.1 E-7 9.6 E-8 1.2 E-7 1.0 E-7 1.6 E-7 1.3 E-7 1.2 E-7 1.0 E-7 1.2 E-7 1.0 E-7 1.1. E-7 8.6 E-8 8.0 E-8 6.5 E-8 6.1 E-8 5.0 E-8 5.8 E-8 4.7 E-8 5.3 E-8 4.5 E-8 4.1 E-8 3.4 E-8 4.5 E-8 3.8 E-8 6.6 E-8 5.5 E-8 6.3 E-8 5.3 E-8 5.1 E-8 4.3 E-8 7.1 E-8 6.0 E-8 4.5 E-8 3.5 E-8 3.2 E4 5.0 E-8 4.4 E-8 8.1 E-8 7.1 E-8 8.9 E-8 7.9 E-8 1.1 E-7 1.0 E-7 8.5 E-8 7.5 8.3 E-8 7.3 E-8 7.4 E-8 6.4 E-8 5.4 E-8 4.6 E-8 4.0 E-8 3.4 E-8 3.8 E-8 3.2 E-8 3.7 E-8 3.2 E-8 2.7 E-8 2.3 E-8 3.2 E-8 2.8 Z-8 4.6 E-8 4.0 Z-8 Di ection wind blowers into

~

~

~ ~

~

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~

~

~

~

~

~

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s g

a

~

~

~

~

4

~

~

TABLE 8 DISPERSXON PARAMETER (D Q)

FOR LONG TERM RELEASES

> 500 HR/YR OR > 125 HR/QTR Ground Vent Distance to the control location, in miles Sector*

0-0.5 0.5-1.0 l., 0.-1, 5 1, 5.-2..0.

2. 0-2. 5.

2. 5-3..0 3.0.-3.5 3.5>>4.0 4.0-4.5 4.5-5.0 NNE N&

ENlE ESE SSE SSW SH hSN 2,0 E>>7 1.8 E-7 2.5 E-7 2.1 E-7 2.5 E-7 2.2 E-7 1.8 E-7 9.8 E-8 6.8 E-8 6.7 E-8

7. 6 E-8 9.9 E-8 1.1 E-7 8.9 E-8 7.0 E-8 l.2 E-7 3.7 E-8 3.4 E-8 4.5 E-8 3.9 E-8

4. 6 =E<<8 4.1 E-8 3.7 E-8 1.8 E-8 1.3 E-8 1.2 E-8 1.4 E-8 1.8 E-8 2.0 E-8 1.6 E-8 1.3 E-8 1.2 E-8 1.2 E-8 5.0 E-9 2.3 E-9 1.4 E-9 1.1 E-8 4.5 E-9 2.1 E-9 1.5 E-8 6.1 E-9 2.8 E-9 1.3 1.7'.3 E-8 5.3 E>>9 1.5 E-8 6.2 E-9 2.4 E-9 2.8 E-9 1.5 E-'9 1.7 E-9 1.3 E-8 5.5 E-9 2.5 E-9 l,l E-8 4.5 E-9 2.1 E-9 6.0 E-9 2.4 E-9 1.1 E-9 4.2 E-9 1.7 E-9 7.7 E-10 1

6 1..3 6;8 4.8 E-9'-10 E-10

4. 3 E-9 l. 7 E-9 7.1 E-9 1.9 E-9 7.9 E-10 1.3 E-9'.

9 E-10 8.1 E-10 4.1 E-9 1.7 E-9 7.6 E-10 4.7 E-10 4.7 E-9 1.9 E-9 8.6 E-10 5.5 E-10 6.1 E-9 1.5 E-9 1.1 E-9 6.9 E-10 6.7 E-9 2.7 E-9 1.2 E-9 7.5 E-10 5.4 E-9 2.2 E-9 1.0 E-9 6.3 E-10 9.7 E-lo 9.0 E-10 1.1 E-,9 1.0 E-9 1.2 E-9.

1.1 E-9'.

0 E-10 4.8 E-10 3.3 E-10 3.3 E-10 3.8 E-10 4.9 E-10 5.4 E-10 4.3 E-10 3.4 E-10 5.7 E-10 7.6 E<<10 5.5 E-10 6.9 E-'10 5.0 E-10 9.2 E-*10 6.9 E-10 8.0 E-10 6.0 E<<10 9.4 E-10 8.4 E-10 7.0 E-10 6.3 E-10 6.9 E-10 S.l E-10 3.7 E-10 2.7 E-10 2.6 E-10 1.9 E-10 2.5 E-10 1.8 E-10 2.9 E"10 2.1 E-10 3.7 E-10 2.8 E-10 4.1 E-10 3.0 E-10 3.3 E-10 2.5 E-10 2.6 E-10 2.0 E-10 4.4 E-10 3.2 E-10 4.7 E-10 4.3 E-lo 5.8 E-0 S.O E

S.8 E-10 5.2 E-10 4.3 E-10 2.3 E-10 1.6 E-10 1.5 E-10 1.7 E-10 2.3 E-10

2. 5.E-10 2.1 Z 0

1.6 E-0 2.7 E-10

1 I

TABLE 9 PATHWAY DOSE FACTORS DUE TO RADIONUCLIDES OTHER THAN NOBLE GASES""

Radio-nuclide H-3 CR-51 MN-54 FE-59 CO-58 CO-60 ZN-65 SR-89 SR-90 ZR-95 I-131 I-133 CS-134 CS-136 CS-137 BA-140 CE-141 Inhalation Pathway x

(mrem/yr3

'/

/

1.12E 03 1.70E 04 1.57E 06 1.27E 06 1.10E 06 7.06E 06 9.94E 05 2.15E 06 1.01E 08 2.23E 06 1.62E 07 3.84E 06 1.01E 06 1.71E 05 9.05E 05 1.74E 06 5.43E 05 Meat Pathway (m

mrem/yr

'/

/

2.33E 02

'.98E 05 7.60E 06 6.49E 08 9.49E 07 3.61E 08 1.05E 09 4.89E 08 1.01E 10 6.09E 08 2.60E 09 6.45E 01 1.42E 09 5.06E 07 1.27E.09 5.00E 07 1.45E 07

.Ground Plane Pat/way 2

(m mrem/yr

/

0.

5.31E 06 1.56E 09 3.09E 08 4.27E 08 2.44E 10 8.28E 08 2.42E 04 0.

2.73E 08.

1.01E 07 1.43E 06 7.70E 09 11.64E 08 1.15E 10 2.26E 07 1.48E 07 Cow-Milk-Infant Pathway i

(m.mrem/yr

/'.38E 03 5.75E 06 3.70E 07 4.01E 08

'.01E 07 2.25E 08 1.99E 10 1.28E 10 1.19E 10 8.76E 05 4.95E 11 4.62E 09 6.37E 10 6.61E 09 5.75E 10 2.75E 08 1.43E 07 Leafy Vegetables Pat/way I.

(m mrem/yr)

/'

2.47E 02 1.63E 06 5.38E 07 1.10E 08 4.55E 07 1.54E 08 2.24E 08 5.39E 09 9.85E 10 1.13E 08 2.08E 10 3.88E 08 1.96E 09 1.60E 08 1.80E 09 2.03E 08 8.99E 07

<Additional dose factors for isotopes not included in Table 9 may be calculated using the methodology described in NUREG-0133, Section 5.3.1 (reference 2)

~

IZ.

Pre aration of S ecial Re ort to Demonstrate Com liance with Environmental Radiation Protection Standards Ginna Technical Specification 3.9.2.4.a requires the preparation and submittal of a Special Report to the Commission when calculated effluent release doses exceed twice the limits of Specifications 3.9.1.2.a, 3.9.2.2.a or 3.9.2.2.b.

In addition, subsequent releases are to be limited so that the dose or dose commitment to a real individual from all uranium fuel cycle sources is limited to < 25 mrem to the total body or any organ (except the thyroid, which is limited to < 75 mrem) for the calendar year that includes 'the release(s) in the Special Report.

This includes the dose contributions from the calendar quarter in which the limits were exceeded and the subsequent calendar quarters within the current calendar year.

The following general guidelines are presented, for preparation of the Special Report:

1)

The maximally exposed real member of the public will generally be the same individual considered in the Technical Specification; 2)

Dose contributions to the maximally exposed individual need only be considered to be those resulting from the Ginna plant itself.

All other uranium fuel cycle facilities or operations are of sufficient distance to contribute a negligible portion of the individual's dose.

3)

For determining the total dose to the maximally exposed individual from the major gaseous and liquid effluent pathways and from direct, radiation, dose evaluation techniques used in preparing the Special Report may be those described in this manual or other applicable methods where appropriate.

4)

The contribution from direct radiation may be estimated by effluent dispersion modelling or calculated from the results of the environmental monitoring program for direct radiation.

0 4.

j'

References 1.

R.E.

Ginna Nuclear Power Plant Unit No.

1, A

endix A to Provisional 0 eratin License No. DPR-18 Technical S ecifications, Rochester Gas and Electric Corporation, Docket No. 50-244.

2.

USNRC, Pre aration of Radiolo ical Effluent Technical S ecifications for Nuclear Power Plants, NUREG-0133 (October, 1978)

~

3.

USNRC, Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Pur ose of Evaluatin Com liance with 10 CFR Part 50 A

endix I, Regulatory Guide 1.109, Revision 1 (October, 1977).

4.

R. E. Ginna Nuclear Power Plant, Environmental Re ort, Appendix B

(August, 1972).

5.

R.E.-Ginna Nuclear Power Plant, Calculations to Demonstrate Com liance

'ith the Desi n Ob ectives of 10 CFR Part 50 A

endix I.

Rochester Gas and Electric Corporation, (June, 1976) 6.

USNRC, Methods for Estimatin Atmos heric Trans ort and Dis ersion of Gaseous Effluents in Routine Releases from Li ht-Water-Cooled Reactors, Regulatory Guide l.ill, Revision 1 (July, 1977).

7.

8.

R. E. Ginna Nuclear Power Plant, Incident Evaluation Ginna Steam Generator Tube Failure Incident Januar 25 1982, Rochester Gas and Electric Corporation, (April 12, 1982).

Pelletier, C. A. et. al., Sources of Radioiodine at Pressurized Water

Reactors, EPRI NP-939 (November, 1978).

e q-.

o e i"'"