ML20081B787

Rev 0 to, Vermont Yankee Nuclear Power Station Offsite Dose Calculation Manual

ML20081B787
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Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	03/05/1984
From:	VERMONT YANKEE NUCLEAR POWER CORP.
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PROC-840305, NUDOCS 8403120005
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{{#Wiki_filter:7 J r / 3 i VERMONT YANKEE NUCLEAR POWER STATION 0FF-SITE DOSE CALCULATION MANUAL a \\ Revision 0. March 1984 t 'l s' r 9 f# g /! e f Yankee Atomic Electric Company Nuclear Services Division 1671 Worcester Road Framingham, Massachusetts 01701 6403120005 840305 PDR ADOCK 05000271 P PDR

REVISION RECORD Revisi g Atti Descrintion 0 March 5, 1984 Initial printing. Reviewed by PORC and approved by management. I I i t i. s I -

LIST OF AFFECTED PAGES Changes, deletions or additions in the most recent revision are indicated by a bar in the margin or by a dot near the page number if the entire page is affected. h Revision All 0 i i -111-

DISCLAINER OF RESPONSIBillTY This document was prepared by Yankee Atomic Electric Company on behalf of Vermont Yankee Nuclear Power Company. This document is believed to be completely true and accurate to the best of our knowledge and information. It is authorized for use specifically by Yankee Atomic Electric Company, Vermont Yankee Nuclear Power Company and/or the appropriate subdivisions within the Nuclear Regulatory Commission only. With regard to any unauthorized use whatsoever, Yankee Atomic Electric Company, Vermont Yankee Nuclear Power Company and their officers, directors, egents and employees assur.e no liability nor make any warranty or representation with respect to the contents of this document or to its accuracy or completeness. -iv- -. ~ -

ABSTRACT The VYNPS 00CM (Vermont Yankee Nuclear Power Station Off-Site Dose Calculation Manual) contains the approved methods to estimate the doses and radionuclide concentrations occurring beyond the boundaries of the plant caused by normal plant operation. With initial approval by the U.S. Nuclear Regulatory Commission and the VYNPS Plant Management and approval of subsequent revisions by the Plant Management (as per the Technical Specifications) this ODCM is suitable to show compliance where referred to by the Plant Technical Specifications. Sufficient documentation of each method is provide 1 to allow an experienced Health Physicist to understand and regenerate the methods with few references to other material. Most of the methods are presented at two levels. The first, Method I, is typically a linear equation which provides an upper bound and the second, Method II, is an in-depth analysis which can provide more realistic estimates. l l _y_

TABLE OF CONTENTS Paae REVISION REC0RD.................................................. 11 LIST OF EFFECTIVE PAGES.......................................... iii DISCLAIMER OF RESPONSIBILITY..................................... iv A8STRACT......................................................... v LIST OF FIGURES.................................................. viii LIST OF TABLES................................................... ix

1.0 INTRODUCTION

1-1 1.1 Summary of Methods, Dose Factors, Limits, Constants, Variables and Definit 1cns.................................. 1-2 2.0 METHOD TO CALCULATE 0FF-SITE LIQUID CONCENTRATIONS............... 2-1 ENG NG 2.1 Method to Determine F and C 2-1 2.2 MethodtoDetermineRddionuclide........................... Concentration for Each Liquid Effluent Pathway........................... 2-2 1 2.2.1 Sample Tanks Pathway............................... 2-2 2.2.2 Service Water Pathway.............................. 2-3 2.2.3 Circulating Water Pathway.......................... 2-3 3.0 0FF-SITE DOSE CALCULATION ME1H0DS................................ 3-1 3.1 Introductory Concepts...................................... 3-2 3.2 Method to Calculate Total Body Dose from Liquid Releases................................................... 3-5 3.3 Method to Calculate Maximum Organ Dose from Liquid Releases................................................... 3-11 3.4 Method to Calculate the Total Body Dose Rate from Noble Gases................................................ 3-14 3.5 Method to Calculate the Skin Do.te Rate from Noble Gases.... 3-18 3.6 Method to Calculate the Critical 0,'gan Dose Rate from Iodines, Tritium and Particulates with T1/2 Greater Than 8 Days.............................. 3-22 3.7 Method to Calculate the Gamma Air Dose from Noble Gases.... 3-26 3.8 Method to Calculate the Beta Air Dose from Noble Gases..... 3-29 i 3.9 Method to Calculate the Critical Organ Dose from Tritium, lodines and Particulates................................... 3-32 -vi-

TABLE OF CONTEN15 (Conting i), Page 3.10 Receptor Points and Annual Average Atmospheric Dispersicn Factors for Important Exposure Pathways......... 3-38 3.11 Method to Calculate Direct Dose from Plant Operation....... 3-42 3.12 Cumulative Doses........................................... 3-44 4.0 ENVIRONMENTAL MONITORING PR0 GRAM................................. 4-1 5.0 SETPOINT DETERMINATIONS.......................................... 5-1 5.1 Liquid Effluent Instrumentation Setpoints.................. 5-2 5.2 Gaseous Effluent Instrumentation Setpoints................. 5-9 6.0 LIQUID AND GASEOUS EFFLUENT STREAMS, RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS................................... 6-1 5.1 In-Plant Liquid Effluent Pathways.......................... 6-1 6.2 In-Plant Gaseous Effluent Pathways......................... 6-3 REFERENCES....................................................... R-1 i l L. -v11-

LIST OF FIGURES Number Title Page 4-1 Environmental Sampling Locations in Close Proximity to Plant 4-4 4-2 Environmental Sampling Locations Within 5 km of Plant 4-5 4-3 Environmental Sampling Locations Greater Than 5 km from Plant 4-6 4-4 TLD Locations in Close Proximity to Plant 4-7 4-5 TLD Locations Within 5 km of Plant 4-8 4-6 TLD Locations Greater Than 5 km from Plant 4-9 6-1 Liquid Effluent Streams, Radiation Monitors and Radwaste Treatment System at Vermont Yankee 6-9 6-2 Gaseous Effluint Streams, Radiation Monitors and Radwaste Treatment System at Vermont Yankee 6-10 -viii-

LIST OF TABLES Number Title Pace ?.1-1 Summary of Radiological Effluent Technical Specifications and plementing Equations 1-3 1.1 Summary of Methods to Calculate Unrestricted Area Liquid Concentrations 1-6 1.1-3 Summary of Methods to Calculate Off-Site Doses from Liquid Releases 1-7 1.1-4 Summary of Methods to Calculate Dose Rates 1-8 1.1-5 Summary of Methods to Calculate Doses to Air f rom Noble Gases 1-9 1.1-6 Summary of Methods to Calculate Dose to an Individual f rom Tritium, Iodine and Particulates 1-10 1.1-7 Summary of Methods for Setpoint Determinations 1-11 1.1-8 Summary of Variables 1-12 1.1-9 Definition of Terms 1-16 1.1-10 Dose Factors Specific for Vermont Yankee 1-18 1.1-11 Dose Factors Specific for Vermont Yankee for Liquid Releases 1-19 1.1-12 Dose and Dose Rate Factors Specific for Vermont Yankee for Tritium, Iodine and Particulate Releases 1-20 3.2-1 Environmental Parameters for Liquid Effluents at Vermont Yankee 3-9 3.2-2 Usage Factors for Various Liquid Pathways at Vermont Yankee 3-10

3. 9-1 Environmental Parameters for Gaseous Effluents at Vermont Yankee 3-36 3.9-2 Usage Factors for Various Gaseous Pathways at l

Vermont Yar.kee 3-37 l. 3.10-1 Vermont Yankee Dilution Factors 3-41 4.1 Radiological Environmental Monitoring Stations 4-2 -ix-

1.0 INTRODUCTION

This ODCM (Off-Site Dose Calculation Manual) provides formal and approved methods for the calculation of of f-site concentratien, off-site doses and effluent monitor setpoints in order to comply with the Vermont Yankee Technical Specifications 3.8/4.8 and 3.9/4.9, hereafter referred to as the Radiological Effluent. Technical Specifications. The ODCM forms the basis for plant procedures and is designed for use by the procedure writer. In addition, the ODCM will be useful to the writer of periodic reports required by the NRC on the dose consequences of plant operation. The initial version of this ODCM is approved by the NRC. The NRC requires notification of revision, but not prior approval. Revisions must be reviewed by PORC and approved by management (see Technical Specifications 6.2.A.6.a and 6.13). I The methods contained herein follow accepted NRC guidance, as of this i.riting, unless otherwise noted in the text. The basis for each method is sufficiently documented to allow regeneration of the methods by an experienced Health Physicist. t 1 -1

l 1.1 Summary ef, Methods. Dose Factors. Limits. Constants. Variables and Definitions This section summarizes the methods for the user. The first-time user should read Chapters 2 inrough 6. The concentration and setpoint methods are documented in Table 1.1-2 through Table 1.1-7, as well as the Method I Dose equations. Where more accurate dose calculations are needed use the Method II .or the appropriate dose as described in Sections 3.2 through 3.9 and 3.11. The dose factors used in the equations are in Tables 1.1-10 through 1.1-12 and the Regulatory Limits are summarized in Table 1.1-1. The variables and special definitions used in this ODCM are in Tables 1,i-8 and 1.1-9. 1-2 .. _ ~

Table 1.1-1 i Summary of Radiological Effluent Technical Specifications and Implementing Equations l -Technical (1) SDecification Category Method _ Limit 3.8.A.1 Liquid Effluent . Total Fraction of Eq. 2-1 5 1.0 Concentration MPC Excluding Noble Gases Total Noble Gas Eq. 2-2 1 2 x 10-4 pC1/cc Concentration 3.8.8.1 Liquid Effluent Total Body Dose Eq. 3-1 5 1.5 mrem in a qtr. Dose 1 3.0 mrem in a yr. Organ Dcse Eq. 3-3 1 5 mrem in a qtr. 1 10 mrem in a yr. 3.8.C.1 Liquid Radwaste Total Body Dose Eq. 3-1 $ 0.06 mrem in a mo. Treatment Operability Organ Dose Eq. 3-3 < 0.2 mrem in a mo. 3.8.E.1 Gaseous Effluents Total Body Dose Rate Eq. 3-5 5 500 mres/yr. Dose Rate from Noble Gases Skin Dose Rate from Eq. 3-7 5 3000 mres/yr. Noble Gases Organ Dose Rate from Eq. 3-16 5 1500 mrem /yr. Iodines Tritium and Particulates with 11/2 > 8 Days 1-3

Table 1.1-1 (continued) Summary of Radiological Effluent Technical Specifications and Implementing Equations Technical (1) Specification Category Method Limit 3.8.F.1 Gaseous Effluents Gamma Air Dose from Eq. 3-21 5 5 mrad in a qtr. Dose from Noble Noble Geses Gases s 10 mrad in a yr. Beta Air Dose from Eq. 3-23 110 mrad in a qtr. Noble Gases 5 20 mrad in a yr. 3.8.G.1 Gaseous Effluents Organ Dose from Eq. 3-25 5 7.5 mrem in a qtr. Dose from lodines, Tritium and

lodines, Particulates with 5 15 mrem in a yr.

Tritium, and Tl/2 > 8 Days Particulates 3.8.I.1 Venttiation Organ Dose Eq. 3-25 5 0.3 mrem in a mo. Exhaust Treatment ) 3.8.M.1 Total Dose (from Total Body Dose Footnote (2). 5 25 mrem in a yr. All Sources) Organ Dose 1 25 mrem.in a yr. l Inyroid Dose 1 75 mrem in a yr.

Table 1.1-1 (continued) Summary of Radiological Effluent Technical Spectilcations and Implementing Equations i fechnical (i) 1 Sr,ecification _ Category Method _ Limit j 3.9.A.1 Liquid Effluent Monitor Setpoint Liquid Radweste Alarm Setpoint Eq. 5-1 T.S. 3.8.A.1 l Discharge Monitor 3.9.B.1 Gaseous Effluent Monitor Setpoint Plant Stack and Alarm / Trip Setpoint Eq. 5-9 T.S. 3.8.E.la A0G Offgas System for Total Body Dose (Total Body) Noble Gas Rate Activity Monitors Alarm / Trip Setpoint Eq. 5-10 1.S. 3.8.E.la for Skin Dose Rate (Skin) SJAE Noble Gas Alarm Setpoint Eq. 5-21 1.S. 3.8.K.1 Activity Monitors 1 (1)- More accurate methods may be available (see subsequent chapters). 1 l (2) Technical Specification 3.8.M.2 requires this evaluation only if twice the limit of equations 3-1, 3-3, 3-21, 3-23 or 3-25 is reached. If this occurs a Method II calculation, using actual meterology and identified pathways for a real individual, shall be made. 1 1-5 \\

Table 1.1-2 Sununary of Methods to Calculate Unrestricted Area Liquid Concentrations Equation Number Category Ea.t1on 2-1 Total Fraction of MPC in ENG i MPC, I I F = Liquids, Except Noble Gases j 2-2 Total Activity of Dissolved NG (,uCi) = EC NG C j ,) g and Entrained Noble Gases from a'11 Station Sources 1 2E-04 i f 1-6

Table 1.1-3 Sumary of Methods to Calculate Off-Site Doses from Liquid Releases Equation Mumber Category Ecuation 3-1 Total Body Dtb(mrem) = 1, Qg DFL ith Dose 1 3-3 Maximum 0,(arem) = E Og DFL, g Organ Dose i 1-7

Table 1.1-4 Summary of Methods to Calculate Dose Rate: Equation Number Category Eauation 3-5 Total Body Dose Rate g gerem) = 0.50 E h DFB 1 1 from Noble Gases tb yr 3 3-7 Skin Dose Rate g Imrem) " DF, from Noble Gases skin yr i g 3-16 Critical Organ Dose Rate f.om Iodines, Egg ( r)" i DFGjc, Tritium i and Particulates with T 1/2 Greater Than Eight Days I l i' .1-8

m Table 1.1-5 i Summary of Methods to Calculate Doses to Air from Noble Gases Equation Number Category Ecuation 3-21 Gamme Dose to Air T T Dair (mrad) = 0.023 }[ Q DF i i from Noble Gases j 3-23 Beta Dose to Air from Noble Gases air (mrad) = 0.02 ][ Q DF D i i g ie t I i 1 } l t f l 1-9 l l

i Table 1.1-6 Summary of Methods to Calculate Dose to an Individual from Tritium, Iodine and Particulates Equation Number Category Eauation 3-25 Dose to Critical Dco (mrem) = EQ DFG i ico Organ from Iodines, j Tritium and Particulates 3-27 Direct Dose Dd (mrem) = 1.29E-06 E l 2 1-10

Table 1.1-7 Sunenary of Methods for Setpoint Determinations Equation Number Category Ecuation 5-1 Liquid Effluents: Liquid Radwaste DF Discharge Monitor setpoint ICPSI " 0F I 7 "I R min i (17/350) Gaseous Effluents: Plant Stack (RR-108-1A, RR-108-18) and A0G Offgas System (3127, 3128) Noble Gas Activity Monitors 5-9 Total Body Rtb (cpm) = 1000 SgiDF8c 5-10 Skin g in (cpm) = 3000 Sg E DF' 5-21 SJAE Noble Gas 1 Activity Monitors setpoint (cpm) = 1.6E+05 SgF R (17/150A, 17/1508) 1-11 i

Table 1.1-8 Summary of Variables ytriable Definition Units C = Concentration at point of discharge of pC1/mi dissolved and entrained noble gas "1" in liquid pathways from all station sources C = Total activity of all dissolved and entrained uti noble gases in liquid pathways from all m1 station sources C = Concentration of radionculide "i" at the point di d1 of liquid discharge ml C, = Concentration of radionuclide "1" pC1/cc C = Concentration, exclusive of noble gases, of uti pi radionuclide "1" from tank "p" at point of ml discharge C,9 = Concentration of radionuclide "1" in mixture pCi/ml at the monitor D = Beta dose to air mrad ar D }r = Gamma dose to air wrad a D, = Dose to the ctitical organ mrem D = Direct dose mrem d Dfinite = Gamma dose to air, corrected for finite cloud mrad i: D, = Dose to the maximum organ mrem D = Dose to skin from beta and gamma mrem D = Dose to the total body mrem tb DF = Dilution factor ratio nim m all wa e u n ac r rado DF = ain 1-12 q.

w Table 1.1-8 (continued) Sunenary of Variables Variable Definition Units DFj = Composite skin dose factor mrem-m pCi-yr 3 DFB = Total body gamma dose factnr for nuclide "1" 9 C 3 DFB = Composite total body dose factor c C DFL = Site-specific, total body dose factor for a mrem itb liquid release of nuclide "1" Ci DFL = Site-specific, maximum organ dose factor for a mrem I"' liquid release of nuclide "1" Ci DFG = Site-specific, critical organ dose factor for a mrem IC8 gaseous release of nuclide "i" C 1 DFG' = Site-specific, critical organ dose rate factor mrem-sec co for a gaseous release of nuclide "1" pC1-yr 3 DFS , Beta skin dose factor for nuclide "1" g -5' DFj = Combined skin dose factor for nuclide "1" -r "r d DF} = Gamma air dose factor for nuclide "1" C_ "r d DFf = Beta air dose factor for nuclide "1" C_ D' = Predicted annual average critical organ yr C dose rate due to iodines and particulates I = Predicted annual average skin dose rate skin r due to noble gases "I'" 3 = Predicted annual average total body dose tb yr rate due to noble gases 1-13

Table 1.1-8 (continued) Summary of Variables Variable Definition Units D/0 = Deposition factor for dry deposition of sec elemental radioiodines and other particulates ,2 E = Gross electric output over the period of MW h interest F = Flow rate out of discharge canal gpm d F, = Flow rate past liquid radwaste monitor gpm F = Flow rate past gaseous radwaste inonttor cc sec F = Total fraction of MPC in liquid pathways fraction (excluding noble gases) MPC = Maximum permissible concentration for di_ g radionuclide "i" (10CFR20, Appendix B, cc Table 2 Column 2) Q = Release for radionuclide "1" curies g h = Release rate for radionuclide "1" pcuries/sec g R = Liquid monitor response for the limiting cps setpoint concentration at the point of discharge R = Response of the noble gas monitor at the cpm skin limiting skin dose rate R = Response of the noble gas monitor to cpm tb limiting total body dose rate S = Shielding factor Ratio p S = Detector counting efficiency from the most CDm mR/hr or I recent gas monitor calibration pCi/cc pC1/cc S' = Detector counting efficiency for noble com mR/hr i r I gas "1" pCi/cc pCi/cc S = Detector counting efficiency from the most cos j recent liquid monitor calibration yC1/ml 1-14 D'

Table 1.1-8 (continued) Sununary of Variables -Variable Definition _ Units S = Detector counting efficiency for cos jg radionuclioe "1" pC1/ml X/Q = Annual average undepleted atmospheric 3 dispersion factor m [X/Q]T = Effective average gamma atmospheric 3 dispersion factor m 1-15

Table 1.1-9 Definition of lerms Annual Averaae Dose - an estimate of dose assuming a one-year exposure period. Annual Average Dose Rate - the dose rate assuming a one-year exposure period. Annual Dose - the fif ty-year dose connitment frem one-year's worth of plant l operation. Composite Sample - generally means a sample collected continuously during the period of release. Conservative Increment in Annual Averaae Dose - a calculated quantity which usually provides an upper bound to the dose from a given activity. Critical Receptor - a hypothetical individual whose location and behavior cause him or her to receive a dose greater than any possible real individual. Dose - the fif ty-year dose commitment measured in mrem to tissue or mrad to air. Dose Pate - the rate for a specific averaging time (i.e., exposcJre period) of dose accumulation. Exposure Period - period of uptake of contaminated air, water or food; or period of exposure to direct radiation. Synonymous with averaging time in most calculations. Limits - the regulations in 10CFR20. Liould Radwaste Treatment System - the system as shown in Figure 6-1. LLD - the " Lower Limit of Detection" as defined in Note a of Table 4.8-1 in the Technical Specifications. I i 1-16

b Table 1.1-9 (continued) Definition of Terms Obiective - the regulations in 10CFR50, Appendix 1. Predicted Annual Average Dose Rate - the annual average dose calculated using the current monitoring system reading and the annual average dilution factors. itandards - the regulations in 40CFR190. l 1-17

s ~, Taple 1.'1-10 'oseFactorsSpechicforVermontYankee 'D . for Noble Gas Releases Gamma Total Body 8 eta Skin Combined Skia Beta Air Gama Air (. Dose Factor Dose Factor Dose Factor Dose Factor Dose Factor 3 3 3 0F8 (mrad-m ) DF}(mrad-m I $(mrem-a)DFj(mrem-sec) aree-m Radionuclide OF84 (mci-vr ) DFS i DCi-vr DCi-vr oCi-vr uti-vr s Ar-41 8.84E-03* 2.69E-03 6.90E-03 3.28E-03 9.30E-03 1.08E-05 2.88E-04 1.93E-05 Kr-83m 7.56E-08 Kr-85m-1.17E-03 1.46E-03

1. 61 E-03 1.97E-03 1.23E-03 8.54E-04 1.95E-03 1.72E-05 Kr-85 1.61E-05 1.34E-03 s.

Kr-87 5.92E-03 9.73E-03 9.59E-03 1.03F-02 6.17E-03 Kr-88 1.47E-02 2.37E-03 1.00E-02 2.93E-03 1.52E-02 Kr-89 1.66E-02 1.01F-02 1.61E-02 1.06E-02 1.73E-02 Kr-90 1.56E-02 7.29E-03 1.37E-02 7.83E-03 1.63E-02 Xe-131m C.15E-05 4.76E-04 3'.87E-04 1.11E-03 1.55E-04 Xe-133m 2.51E-04 9.94E-04 ' '8.09E-04 l'.48E-03 3.27E-04 Xe-133 2.94E-04 3.06E-04 3.90E-04 1.05E-03 3.53E-04 Xe-135m 3.12E-03 7.11E-04

• 6.36E-04 7.39E-04 3.36E-03 i

Xe-135 1.81E-03 1.86E-03$ - 2.25E-03 2.46E-03 1.92E-03 X2-137 1.42E-03 1.22E-02 8.53E-03 1.27E-02

1. 51 E-03 X7-138 8.83E-03 4.13E-03 7.76E-03 4.75E-03

9. 21 E-03 0 8.84E-03 = 8.84 x 10-3 1-15

Table 1.1-1_1 Dose Factors Specific for Vermont Yankee for Liquid Releases Total Body Maximum Organ Dose Factor Dose Factor DFL Imrem) DFL,,(mrem) Radionuclide ith Ci 4 Ci H-3 1.SiE-06 1.99E-06 Na-24 1.33E-03 1.33E-03 Cr-51 1.20E-05 2.75E-03 Mn-54 7.88E-03 1.18E-01 Mn-56 3.29E-07 2.11E-04 Fe-59 9.07E-03 7.06E-02 Co -58 1.89E-03 1.58E-02 Co-60 5.40E-03 4.24E-02 Zn-65 3.14E-01 6.48E-01 Sr-89 7.73E-03 2.71E-01 Sr-90 1.18E+00 4.80E+00 Zr-95 5.14E-07 2.13E-03 No-99 1.78E-04 1.73E-03 Tc-99m 3.03E-06 1.43E-04 Sb-124 2.66E-05 1.66E-03 I-131 9.90E-04 5.66E-01 1-132 4.83E-08 4.87E-06 I-133 1.29E-04 6.36E-02 1-135 1.30E-05 2.43E-03 Cs-134 5.10E+00 6.32E+00 Cs-137 3.01E+00 4.79E+00 Ba-140 1.27E-04 3.34E-03 Ce-141 1.99E-08 5.00E-04 W-187

4. 72E- 04 3.56E-01 1-19 I

Table 1.1-12 Dose and Dose Rate Factors Specific for Vermont Yankee

for, Iodines, Tritium and Particulate Releases Critical Organ Critical Organ a

Dose Factor Dose Rate Factor jc,(mrem) DFG,co Imrem-sec) DFG Ci i vr-uC1 Radionuclide H-3 1.70E-04 5.36E-03 C-14 1.13E-01 3.56E+00 Cr-51 3.36E-04 1.06E-02 Mn-54 4.60E-02 1.45E+00 Fe-59 5.13E-02 1.62E+00 Co-58 0.15E-02 6.77E-01 Co-60 7.15E-01 2.26E+01 J Zn-65 1.13E-01 3.55E+00 Rb-88 8.20E-06 2.59E-04 Sr-89 1.10E+00 3.48E+01 .Sr-90 4.25E+01 1.34E+03 Zr-95 5.53E-02 1.74E+00 Sb-124 8.97E-02 2.83E+00 1-131 8.36E+00 2.64E+02 I-133 1.28E-01 4.04E+00 I-135 1.17E-02 3.69E-01 Cs-154 2.84E+00 8.96E+01 Cs-137 2.68E+00 8.45E+01 Ba-140 2.96E-02 9.34E-01 Ce-141 1.52E-02 4.79E-01 Ce-144 3.54E-01 1.12E+01 5 ~ ~ 1-20

2.0 METHOD TO CALCULATE OFF-SITE LIOUID CONCENTRATIONS Chapter 2 contains the basis for plant procedures that the piant operator reauires to meet Technical Specification 3.8.A.1 which limits the total fraction of MPC in liquid pathways, excluding noble gases, denoted here as F ", at the point of discharge (see Figure 6-1). F" is limited to less than or equal to one, i.e., F < 1. 1 The total concentration of all dissolveu and entrained noble gases at the point of discharge from all station sources, denoted C, is limited to 2E-04 pCi/ml, i.e., Cf12E-04pCi/ml. Evaluation of F " andCfisrequiredcorcurrentwiththe sampling and analysis program in Technical Specification Table 4.8.1. 2.1 Method to Determine F " and C Determine the total fraction of MPC at the point of discharge in liquid pathways (excluding noble gases), denoted F ", and determine the total concentration at the point of discharge of all dissolved and entrained noble gases in liquid pathways from all station sources, denoted C , as follows: .i FENG, y $ 1. (2-1) p i i IuCi/ml) uci/ml and: 2-1

t i C"0~ C 1 2E-04 (2-2) = (yci/ml) (yC1/ml) (yC1/ml) where: 1-F" Total fraction of MPC in liquid pathways, excluding noble = gar,es, at the point of discharge Concentration at point of discharge of radionclude "i", C,j = p except for dissolved and entrained noble gases, f rom any tank or other significant source, p, f rom which a discharge may be made (including the floor drain sample tank, the waste sample tanks, the detergent waste ta d and any other significant source from which a discharge can be made) (uti/ml) Maximum permissible concentration of r;dionuclide "i" except MPCj = for dissolved and entrained noble gases from 10CFR20, Appendix B. Table II, Coluinn 2 (yC1/ml) C" Total concentration at point of discharge of all dissolved = and entrained noble gases in liquid pathways from all station sources (yci/ml) Concentration at point of discharge of dissolved and entrained C = noble gas "1" in liquid pathways from all station sources (pCi/ml)- 2.2 Method to Determine 'Radionucliae Concentration for Each Liquid Effluent Pathway 2.2.1 Sample Tanks Pathways C is determined for each radionuclide above LLD from the activity p in a proportional grab sample of any of the sample tanks and the predicted flow at the point of discharge. Most periodic batch releases are made from the two 10,000-gallon capacity waste sample tanks. These tanks serve to hold all the high purity liquid wastes 'after they have been filtered through the waste collector and processed by ion exchange in the fuel pool and waste demineralizers. Other ~ periodic batch -releases may also come from.the detergent waste tank or the floor drain sample tank. 2-2 l )

The tanks are sampled from the radweste sample sink and the contents analyzed for water quality and radioactivity. If the sample meets all the high purity requirements, the contents of the tank may be re-used in the nuclear system. If the sample does not meet all the high purity requirements, the contents are recycled through the radwaste system cr discharged. Prior to discharge each sample tank is analyzed for tritium, dissolved noble gases and dissolved and suspended gamma emitters. 2.2.2 Service Water Pathway The service water pathway shown on Figure 6-1, flows from the intake structure through the heat exchangers and the discharge structure. Under normal operating conditions, the water in tilis line is not radioactive. For this reason, the service water line is not sampled routinely but it is continuou. ly monitored with the service water discharge monitor (No.17/351). The alarm setpoint on the service water discharge monitor is set at a level which is three times the background of the instrument. The service water is sampled if the monitor is out of service or if the alarm sounds. Under normal operating conditions, the concentration of radionuclides at the point of discharge from the service water effluent pathway will never exceed the maximum permissible concentration in 10CFR20, Appendix B Table II, Column 2. -2.2.3 Circulating Water Pathway The circulating water patnway shown on Figure 6-1, flows from the intake structure through tiie condenser and the discharge structure. Under normal operating conditicns, the water in this line is not radioactive For this reason, the circulating water line is not sampled routinely but it is monitored continuously by the discharge process monitor (No.17/359) located in the discharge structure. 2-3

The alarm setpoint on the discharge process monitor is set at a level which is three times the background of the instrument. The circulating water is sampled if the monitor is out of service of if the alarm sounds. Under normal operating conditions, the concentration of radionuclides at the point of discharge from the circulating water pathway will never exceed the maximum permissible concentration in 10CFR20, Appendix B, Table II, Column 2. 2-4

3.0 0FF-SITE DOSE CALCULATION METHODS Chapter 3 provides the basis for plant procedures required to meet the Radiological Effluent Technical Specifications (hereafter called RETS). A simple, conservative method (called Method I) is listed in Tables 1.1-2 to 1.1-7 for each of the requirements of the RETS. Each of the Method I equations is presented, along with their bases in Sections 3.2 through 3.9 and Section 3.11. In addition, those sections include more sophisticated methods (called Method II) for use when more accurate results are needed. This chapter provides the methods, data, and reference material with which the operator can calculate the needed doses, dose rates and setpoints. i ! 1

3.1 Introductory Concepts The Radivlogical Ef fluent Technical Specifications (RETS) either limit dose, denoted "D", or dose rate, denoted "b". The term " Dose" means the fifty-year dose commitment, measured in mrem to human tissue or mrad to air. The phrases " annual Dose" or " Dose in one year" mean the fif ty-year dose commitment from one year's worth of releases. " Dose in a quarter" similarly means from. one quarter's releases. " Dose Rate" is the dose commitment divided by expo,ure p?riod. For example, an individual who is exposed for one year and receives a 50-year dose commitment of 10 mrem is assumed to have received that dose in the year and is said to have been exposed to a 10 mrem / year Dose Rate. The phrase " annual average Dose Rate" means the Dose Rate calculated for a one year exposure period. The concept of averaging time is critical. Because radioactive decay. is a discrete process (i.e., is not continuous) an " instantaneous dose rate" cannot be defined. This is because the limit of any discrete process fluctuates between infinity and zero as dt apprcaches zero. For regulatory purposes some non-zero averaging time has to be chosen. The typical averaging time used here is either one year or the inherent averaging time of the release rate measuring system. For example, the release rate averaging time is one week if the charcoal filters used to detect lodine are changed every week, or the reading of the Primary Vent Stack nobic gas monitor is used directly (a much shorter averaging ti m is built in). I The " annual average Dose" is a statistical quantity and means the Dose estimated from information which has been averaged for a one-year exposure period. -The quantities AD and D are introduced to provide calculable quantities, related to of f-site dose, which demonstrate compliance with the RETS as well as the Limits, Standards and Objectives on which the RETS are based. 3-2

- ~.. Delta D, denoted AD, is the quantity calculated by the Chapter 3 dose equations. It is the conservative increment in annual average Dose. AD is not the actual dose received by a real individual ho+ it usually provides an upper bound for a given release because of the conse.vative margin built into the dose factors and the selection and definition of critical receptors. AD is convenient because it obviates the bookkeeping required to keep track of cumulative dose to multiple real receptors for each release. The radioisotope specific dose factors in each dose equation represent the greatest Dose to any organ of any age group. (Organ Dose is a function of age because organ mass and intake is a function of age.) The critical receptor is a hypothetical individual whose behavior - in terms of location and intake - results in a Dose which is higner than any real individual could possibly receive. D tilde, denoted D, is the quantity calculated in the Chapter 3 dose rate equations. It is the " predicted annual average Dose Rate" and is calculated using the monitoring system reading and the annual average atmospheric dispersion factor. D predicts the maximum off-site annual dose if the maximum release rate continued for one entire year. This assures that 100FR20.106 limits are always met, with a very large conservative margin. In this regulatory. setting, with this conservative margin, it makes no sense to calculate Dose Rate averaged for the release period. The Dose and Dose Rate Limiting Conditions for Operation (LCO), in the RETS differ by level of action. RETS which are based on the limits in 10CFR20' may require plant shutdown. Conversely, the action required in the RETS that are based on the Objectives and Standards in 10CFR50 and 40CFR190, do not immediately limit plant operation but require a report to the NRC. In general, dose LCOs must be evaluated monthly and dose rate LCOs must be evaluated.in accordance with the sampling and analysis program, The methods i in Chapter 3 are designed entirely to comply with the RETS. On occasion the ~ following Regulations provided further detail: 10CFR20.lc 10CFR20.105 10CFR20.106 10CFR20, Appendix B Table II t 3-3

10CF450.34a 10CFR50.36a 10CFR50, Appendix I 40CFR190.10 The conservative margins required differ by !!egulation. The Limits in 10CFR20 are absolute, so that significant margin must be included in the a Methods; 10CFR50 requires " reasonable assurance that the... objectives are met"; and similarly, 40CFR190 requires " reasonable assurance" that the Standards are met. In Chapter 3, statistical methods are used to establish the conservative ;nargin required by " reasonable assurance". Each of the methods to calculate dose or dose rate are nresented in separate sections of Chapter 3, and are summarized in Tables 1.1-1 to 1.1-7. Each method has two levels of complexity and conservative cargin called Method I and Method II. Method I has the greatest margin and is the simplest; generally a linear equation. hethod II is a complete analysis. Guidance is provided but the aopropriate margin and depth of analysis are determined in each instance. Each section of Chapter 3 reviews applicable Technical Specifications, possi' ale Actions, Surveillance Requirements, the use and limits of validity of Method I, criteria for using Method II, and the base case - which is the basis for Method I and the starting point for Method II analyses. 3-4

3.2 Method to Calculate the Total Body Dose from Liauid Releases Technical Specification 3.8.8.1 limits the total body dose commitment .tc a Member of the Public from radioactive material in liquid effluents to 1.5 arem per quarter and 3 mrem per year. Technical Specification 3.8.C.1 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any month. Technical Specification 3.8.M.1 limits the total body dose commitment to any real member of the public from all station sources (including liquids) to 25 mrem in a year. Dose evaluation is required at least once per month. If the liquid radwaste treatment system is not being used, dose evaluation is required before each release. Use Method I first to calculate the maximum total body dose from a liquid release to the Cor.necticut River as it is simpler to execute and more conservative than Method II. Use Method II if a more accurate calculation of total body dose is 6 needed (i.e., Method I indicates the dose is greater than the limit), or if Nethod I cannot be applied. If the radwaste system is not operating, the total body dose must be estimated prior to a release (Specification 3.8.C.1). To evaluate the total body dose, use Equation 3.1 to estimate the dose from the planned release and add this to the total body dose accumulated from prior releases during the month. f. 3.2.1 Method I The increment in total body dose from e liquid release is: 0 (' O tb 1 itb 1 (mrem) (C1)("[{*)- where: l PFlith Site-specific total body dose factor (mrem /C1) for a liquid = release. It is the highest of the four age groups. See ' Table 1.1.11. 3-5

Total activ-aased for radionuclide "i". (For Qj = strontiums. recent measurement available.) Equation 3-1 can be applied cade. . n,. aing conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), 2. Liquid releases were to the Connecticut River, and 3. Any continuous or batch release over any time period. 3.2.2 Method II If Method I cannot be applied, or if the Method I dose e..aeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, . documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. 3.2.3 Basis for Method I This section serves three purposes: (1) to document that Method I . complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. l Method I may be used to show that the Technical Specifications which l limit off-site total body dose from liquids (3.8.B.1 and 3.8.C.1) have been met for releases over the appropriate periods. These Technical Specifications are based on design objectives and standards in 10CFR and 40CFR. Technical Specification 3.8.8.1 is based on the ALARA design objectives in 10CFR50, Appendix I Subsection II A. Technical Specification 3.8.C.1 is an " appropriate fraction", determined by the NRC, of that design objective (hereafter called the Objective). Technical Specification 3.8.M.1 is based on 3-6

Environmental Standards for Uranium Fuel Cycle in 40CFR190 (hereafter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents. Exceeding the Objective or the Standard does not immediately limit plant operation but requires a report to the NRC within 30 days. In addition, a waiver may be required. This is unlike exceeding 10CFR20 limits which could result in plant shutdown. Method I was developed such that "the actual exposure of an individual... is unlikely to be srbstantially underestimated", (10CFR50, Appendix I). The definition, below, of a single " critical receptor" (a hypothetical individual b. hose behavior results in an unrealistically high . dose) provides part of the conservattye margin to the calculation of total body dose in Method I. Method II allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was beted on a Method II analysis for the critical receptor and ( 1nual averaga ' conditions instead of any real individual. That analysis was called the " base case"; it was then reduced to form Method I. The base case, the method of -reduction, and the assumptions and data used are presented below. The steps performed in the Method I derivation follow. First, in the base case, the dose impact to the critical receptor (in the form of dose factors DFLitb, mrem /01) for a 1 curie release of each radioisotope in liquid effluents was derived. The base case analysis uses the methods, data and assumptions-in Regulatory Guide 1.109 (Equations A-2, A-3, A-7, A-13 and A-16 Reference A)..The only liquid pathway contributing to an individual dose is due to consumption of fish from the Connecticut River. A plant discharge flow rate of 44.6 ft /sec was used'with a mixing ratio of 0.04. Tables 3.2-1 and 3.2-2 outline human consumption and environmental parameters used in the analysis. The resulting, site-specific, total body dose factors appear in Table 1.1.11. For any liquid release, during any period, the increment in annual average total body dose from radionuclide "1" is: 3-7

AD =0 DFL (3-2) 9 (mrem) (C1) ("C ) where: DFlitb Site-specific total body dose factor (mrem /C1) for a liquid = release. It is the highest of the four age groups. See Table 1.1.11. Total activity (Ci) released for radionuclide "1". Qi = Method I is more conservative than Method II in the region of the Technical Specification limits because it is based on the following reduction of the base case. The dose factors DFl used in Method I were chosen from itb the base case to be the highest of the four age groups for that radionuclide. In effect each radionuclide is conservatively represented by its own critical age group. Because of the assumptions about receptors, environment, and radionuclides; and because of the low Objective and Standard, the lack of immediate restriction on plant operation, and the adherence to 10CFR20 concentrations (which limit rublic health consequences) a failure of Method I (i.e., the exposure of a real individual being underestimated) is improbable and the consequences of a failure are minimal. 3-8

1 IAbj e_3.2 -1 i Environmelital Parameters for Liquid Effluents at Vermont Yankee (Derived from Reference A) POTABLE AQUATIC SHORELINE FOOD GROWN WITH CONTAMINAlED WATER VARIABLE WATER FOOD AC1IVITY VEGETABLES LEAFY VEG. MEAT COW MILK GOAT MILK MP Mixing Ratio 0.04 0.04 0.04 0.04 0.04 0.04 0.04 0.04 TP Transit Time (HRS) 12.0 24.0 0.0 0.0 0.0 0.0 0.0 0.0 YV Agricultural (KG/M ) 2.0 2.0 0.70 0.70 0.70 2 l Productivity P Soil Surface (KG/M ) 240.0 240.0 240.0 240.0 240.0 2 Density (L/M /HR) 0.0 0.0 0.0 0.0 0.0 2 i IRR Irrigation Rate TE Crop Exposure (HRS) 1440.0 1440.0 720.0 720.0 720.0 Time TH Holdup Time (HRS) 1440.0 24.0 0.0 48.0 48.0 QAW Water Uptake Rate (L/0) 50.0 60.0 8.0 for Animal QF Feed Uptake Rate (KG/D) 50.0 50.0 6.0 for Animal Location of None Connecticut None None None None None None Critical Individual River 3-9

E e-o a E 8 8 8 8 >s Q O O O <E lE 2 ':w w w Z ^ e-s E O O O O O. C. O. O. J w E E N N O O Z e e-Z w 9) ges=s $ M ^ g E. m e W l>

, x

_J E s O O O O O. O. O. O. go co w E 4H w ~ d, X W< W O O O O f5 03 w 't - tb J 2: w 4h y R m O W E ~ C E >= O O O O p 6.J N O. O. O. O. y W c mL E M C C O O si e w w-ct M ' EO ^ ,;~t - =, E 8 8 8 ~ y s m _i w a -O o w u ~ ~ a w I =i n e H. W E t 8 8 8. 8 2 t,gg W G E M C m O p w b 0 4 G 'A e U E g C >= k 8 8 8 8 w. H w E H O. O O O g g e w E -J 5 = 3 O w' a w E w O O O O 40 N O. O. O. O. ww e a> M C O O O w JM% E >= O O O O G N O. O. O. O. w

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C O O O w M M M C LaJ m C r== no G 3 9 ar= w ( D e J:: C 4 >= U M

3.3 Method to Calculate Maximum Oroan Dose from Liauid Releases Technical Specification 3.8.B.1 limits the maximum organ dose commitment to a Member of the Public from radioactive material in liquid effluents to 5 mrem per quarter and 10 mrem per year. Technical Specification 3.8.C.1 requires liquid radwaste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any month. Technical Specification 3.8.M.1 limits the maximum organ dose commitment to any real membe. of the public from all station sources (including. liquids) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year. Dose evaluation is required at least once per month if releases have occurred. If the liquid radwaste treatment system is not being used, dose evaluation is required before each release. Use Method I first to calculate the neximum organ dose from a liquid release to the Connecticut River as it is simpler to execute and more conservative than Method II. Use Method II if a more accurate calculation of organ dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied. If the radwaste system is not operating, the maximum organ oose must be estimated prior to a release (Specification 3.8.C.1). To evaluate the maximum organ dose, use Equation 3-3 to estimate the dose from the planned release and add this to the maximum organ dose accumulated from prior releases during the month. 3.3.1 Method I The increment in maximum organ dose from a liquid release is: O "EO DFL (3-3) mo j ISO i 3-11

where: Site-specific maximum organ dose factor (mrem /C1) for a DFLj,o = liquid release. It is the highest of the four age groups. See Table 1.1.11. Qi Total activity (C1) released for radionuclide *1". (For = strontlums, use the most recent measurement available.) Equation 3-3 can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), 2. Liquid releases were to the Connecticut River, and 3. Any continuous or batch release over any time period. 3.3.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method -II analysis. 3.3.3 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calculate neximum organ dose parallel the total body dose methods (see Section 3.2.3). Only the differences are presented here. 3-12

For each radionuclide, a dose factor (mrem /C1) was determined for each of seven organs and four age groups. The largest of these was chosen to be the maximum organ dose factor (DFL, ) for that radionuclide. g For any liquid release, during any period, the increment in annual average dose from radionuclide "1" to the maximum organ is: AD, =Q DFL,9 (3-4) g g (mrem) (C1) ("C I where: i Site-specific maximum organ dose factor (mrem /Ci) for a DFl no = liquid release. See Table 1.1.11. Total activity (Ci) released for radionuclide "1". Qj = Because of the assumptions about receptors, environment, and radionuclides; and because of the low Objective and Standard, the lack of immediate restriction on plant operation, and the adherence to 10CFR20 concentrations-(which limit public health consequences) a failure of Method I (i.e., the exposure of a real individu61 being underestimated) is improbable and the consequences of a failure are minimal. 3-13

3.4 _ Method to Calculate the Total Body Dose Rate From Noble Gases Technical Specification 3.8.E.1 limits the dose rate to the total body from noble gases at any location at or beyond the site h undary to 500 mrem / year. The Technical Specification limits peak release rates by limiting the annual average dose rate that is icted from continued release at the peak rate. The " Predicted Annual Average Total Body Dose Rate", denoted D tb' is the highest total body dose received by anyone off-site in a year if the peak release rate reading were to continue for a year. The peak release rate averaging time is commensurate with the rate constant of the measuring system. By limiting U to 500 mrem / year, we assure that the total body dose tb accrued in any one year by any member of the general public is much less than 500 mrem. The Specificatien requires continuous surveillance. If this Technical Specification is eFceeded and operator actfin Cannot be taken tG reduce the predicted average total body dose rate, then shutdown may be required. Use Method I first to calculate the Predicted Annual Average Total Body Dose Rate from the peak release rate via the plant stack. Method I applies at all release rates. Use Method II if a more accurate calculation of I is desired by the tb plant (i.e., use of actual n;eteorology) or if Method I cannot be applied. 3.4.1 Method I The Total Body Dose Rate due to noble gases can be determined as follows: Eh DFB (3-5) U 0.50 = g tb i 3 rCi-sec uCi Imrem) I mrem-m ) 3 sec C1-yr yr pci-m 3-14

where: Qj Peak release rate that is averaged by the time constant of the = measuring system (uC1/sec) for cach radionuclide, "i", shown in Table 1.1-10. DF81 Total body gamma dose factor (see Table 1.1-10). Equation 3-5 can be cpplied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), and 2. Noble gas releases via the plant stack to the atmosphere. 3.4.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in ' Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. 3.4.3 Basis for Method I This section serves four purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to define the word " rate" as used in the Technical Specification, (3) to provide background and training information to Method I users, and (4) to provide an introductory user's guide -to Method II. Method I may be used to show that the Technical Specification which limits total body dose rste from noble gases released to the atmosphere (Technical Specification 3.8.E.1) has been met for the peak noble gas release rate. 3-15

Method I was derived from Regulatory Guide 1.109 as follows: 5 = 1E+06 S [X/Q]Y EQ DFB (3-6) tb p g g i I I IDCi) I#I I I IuCi) I I are.3 sec mrem-m sec pCi-yr yr yC1 ,3 where: Shielding factor = 0.7. S = p [X/Q)T Maximum annual average gamma atmospheric dispersion factor = 3 7.2E-07 (sec/m ) = h, Release rate of noble gas "1" (yCi/sec). = 3 Gamma total body dose factor, ( ). See Table 1.1-10. DFB g Equation 3-6 reduces to: 0.50 h DFB E = tb 3 i (3-5) 3 Imrem) IDCi-sec) IuCi II I mrem-m 3' sec pCi-yr yr Ci-m Because annual average meteorology is used (see Section 3.10), 5, has tb physical meaning only if Q is averaged over a year, otherwise 5 only has g tb regulatory meaning. The selection of critical receptor, outlined in Section 3.10 is inherent in Method I, as are the maximum expected off-site annual average atmospheric dispersion factors. All noble gases in Table 1.1-10 must be considered, none are deemed insignificant a priori. Method II cannot provide much extra realism because D has little tb physical meaning to begin with. However, should it be needed, the analysis for critical receptor and the annual average atmospheric dispersion factors may be performed on the most recent year's meteorologic data. 3-16

Because of the choice of atmospheric dispersion factors, it is expected that Method I results will exceed Method II calculations 95 percent of the time. Either method provides adequate margin to ensure that the annual average concentrations based on total body dose from 10CFR20.106(d) are not exceeded and that the derived instantaneous release monitor readings are conservative. l k L P 3-17 I

3.5 Method to Calculate the Skin Dose Rate from Noble Gases Technical Specification 3.8.E.1 limits the dose rate to the skin from noble gases at any location at or beyond the site boundary to 3,000 mrem / year. The Technical Specification limits peak release rates by limiting the annual average dose rate that is predicted from continued release at the peak rate. The " Predicted Annual Average Skin Dose Rate", denoted Dsk, is the highest skin dose received by anyone off-site in a year if the peak release rate reading were to continue for a year. The peak release rate averaging time is comensurate with the rate constant of the measuring system. By limitingU to 3,000 mrem / year, we assure that the skin dose accrued in any sk one year by any member of the general public is much less than 3,000 mrem. The Specification requires continuous surveillance. If this Technical Specification is exceeded and operator action cannot be taken to reduce the pr?dicted average skin dose rate, then shutdown may be required. Use Method I first to calculate the Predicted Annual Average Skin Dose Rate from the peak release rate via the plant vent stack. Method I 13pplies at all release rates. Use Method II if a more accurate calculation of D is desired by the sk plant (i.e., use of ariual meteorology) or if Method I cannot be applied. 3.5.1 Method I The Skin Dose Rate due to noble gases is: U DF (3-7) skin " i j i I I Iuti) I I mrem mrem-sec yr sec pCi-yr where: peak release rate that is averaged by the time constant of the Q = measuring system (uC1/sec) for each radionuclide, "1", shown in Table 1.1-10. DFj combined skin dose factor (see Table 1.1-10). = 3-18

Equation 3-7 can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), and 2. Noble gas releases via the plant stack to the atmosphere 3.5.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit cr if a more exact calculation is required, then Method II may be applied. Method Il consists of the models, input data and assumptions in Regulatory Guide 1.109 Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. 3.5.3 Basis For Met. hod I This section serves four purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to define the word " rate" as used in the Technical Specification, (3) to provide background and training information to Method I users, and (4) to provide an introductory user's guide 'to Method II. The methods to calculate skin dose rate parallel the total body dose rate methods in Section 3.4.3. Only the dif ferences are pre _lented here. Method I may be used to show that the Technical Specification which limits skin dose rate from noble gases released to the atmosphere (Technical Specification 3.8.E.1) has been met for the peak noble gas release rate. The annual skin dose limit is 3,000 mrem (from NBS Handbook 69, Reference D, pages 5 and 6, is 30 rem /10), which is the basis for the MPC"

• limits.

The concept of Predicted Annual Average Skin Dose Rate denoted Dsk, is introduced to distinguish it as a regulatory quantity rather than a 3-19

i measurable quantity. It is the skin dose commitment to the critical off-site receptor assuming annual average meteorology and that the release rate reading remains constant over the entire year. Method I was derived from Regulatory Guide 1.109 as follows: D{ir + 3.17E+04 E Qg [X/Q] DFS (3-8) D = 1.11 S g p i 3 I I I ") I") I I IDCi-vr) Ci mrem mrad yr yr Ci-sec F Isec) I I mrem-m 3 pCi-yr m where: 1.11 = Average ratio of tissue to air absorption coef ficients (will convert mrad in air to mrem in tissue. T D = 3.17E+04 [ Q [X/Q] DF (3-9) air g j i 3 Ci ,3 } mrad-m Imrad) IDCi-vr) Iyr) I sec pCi-yr yr Ci-sec D [X/Q]T [X/Q] (3-10) Of 1te = / now 1 1r 3 Imead) I II,3 } mrad sec m Isec) yr yr and Q = 31.54 h (3-11) g g ) Ci-sec) uC1) pCi-yr sec so 5 = 1.11 S M 6 [X/Q F E h DF} (3-12) skin F g 1 ( "y"r ) (#) (#) (pCi) (3) (sec) ("pCi-yr )

• ~"

+ 1E+06 X/Q E h DFS g g 1 3 uQ.1 mrem-m ) (DCi) sec sec pCi-yr Ci ,3 3-20

substituting 7.2E-07 sec/m3 i [X/Q]Y = 6.3E-07 sec/m3 X/Q = Shielding factor = 0.7 Sp = (3-13) + 0.63 1h DFS 0.56 E h DF} gives E = 3 g skin 1 1 Imrem) IDCi-sec-mrem) Iuti) Imrem-m ) IpCi-sec) IuCi) I mrem-m yr 3 sec pci-yr 3 sec pCl-yr pCi-m -mrad nCi-m = E h [0.56 DF} + 0.63 DFS ] (3-14) g j i define DFj=0.56DF}+0.63DFS$ (3-15) then Iskin i DFj (3-7) i mrem) pC1) arem-sec) yr sec pCi-yr Because annual average meteorology is used (see Section 3.10), Dsk, has physical meaning only if Q is averaged over a year, otherwise E only has g sk . regulatory meaning. The selection of critical receptor, outlined in Section 3.10 is inherent in Method I, as it determined the maximum expected off-site atmospheric dispersion factors. All noble gases in Table 1.1-10 must be considered. Because of the choice of atmospher'c dispersion factors, it is expected that Method I results will exceed Method II calculations 95 percent of the time. Either method provides adequate margin to ensure that the annual average concentrations based on the skin dose limit of 3,000 mrem are not exceeded and that the derived instantaneous release monitor readings are conservative. 3-21

--mm

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l.0 52 HE 'j l,f El i i,i [m EM l.8 1.25 1.4 1.6 4 150mm 4 6" b 6t % /A +>;gy> b;i +Q(O i s/s v + e r NJ:=_..

3.6 Method to Calculate the Critical Oraan Dose Rate from Iodines. Tritium and Particulates with T Greater Than 8 Days g i Te':hnical Specification 3.8.E.1 limits the dose rate to any organ from 131, 3 and radionuclides in particulate form with half lives greater 3 H than 8 days to 1500 mrem / year to any organ. The Technical Specification limits peak release rates by limiting the ar.nual average dose rate that is predicted from continued release at the peak rate. The " Predicted Annual Average Critical Organ Dose Rate", denoted Dco, is the highest dose received by anyone off-site in a year if the peak release rate reading were to continue for a year. The peak release rate neraging time is commensurate with the rate constant of the measuring system. By limiting D to 1500 mrem / year, we co assure that the critical organ dose accrued in any one year by any member of the general public is much less than 1500 mrem. The Specification requires continuous surveillance. If this Technical Specification is exceeded and operator action cannot be taken to reduce the predicted average critical organ dose rate, then shutdown may be required. Use Method I first to calculate the Predicted Annual Average Critical Organ Dose Rate from the peak release rate via the plant vent stack. Method I applies at all release rates. Use Method II if a more accurate calculation of S is desired by the co plant (i.e., use of actual meteorology) or if Method I cannot be applieo. 3.6.1 Method I i The Critical Organ Dose Rate can be determined as follows: co i DFGjeg (3-16) i I I IuCi) I I mrem mrem-sec yr sec pCi-yr where: 3-22

h = Peak activity release rate reading cf radionuclide "i" in g pCi/sec. For i = Sr89 or Sr90, use the best estimates (such as most recent measurements). jc, = Site specific critical crgan dose factor (" 'C r ) for a o" sous DFG release. See Table 1.1-12. Equation 3-16 can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), and 2. Tritium, I-131 and particulate releases via the plant vent stack to the atmosphere. 3.6.2 Method II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, thea Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable starting point for a Method II analysis. l 3.6.3 Basis for Method I t This section serves four purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to define the word " rate" as used in the Technical Specification, (3) to provide background and training information to Method I users, and (4) to provide an introductory user's guide to Method II. The methods to calculate critical organ dose rate parallel the total body doss rate methods in Section 3.4.3. Only the differences are presented here, l l 3-23 l

Method I nay be used to show that the Technical Specification which limits organ dose rate from Iodines, Tritium and radionuclides in particulate form with half lives greater than 8 days (hereafter called Iodines and Part*culates or "I+P") released to the atmosphere (Technical Specification 3.8.E.1) has been met for the peak I + F release rates. The annual organ dose limit used here is 1500 mrem (from NBS Handbook 69 Reference D, pages 5 and 6). The concept of Predicted Annual Average Critica' Organ Dose Rate denoted li, is introduced to distinguish it as a regulatory quantity rather than a co measurable quantity. It is the critical organ dose commitment to the critical off-site receptor assuming annual average meteorology and that the release rate reading remains constant over the entire year. The release rate for our purposes is commensurate with the rate constant of the measuring system. The regulatory limits, howeve'r. are met very conservatively by this scheme. Exceeding the Annual Average Critical Organ Dose Rate could result in plant shutdown if the operators cannot take action to reduce the peak release rate. The equation for 5, is derived by modifying Equation 3-25 from c Section 3.9 as follows: Q DM (3 % D, a c g jco (mrem) (Ci) (*C 'I applyingtheconversionfactor,31.54(Ci-sec/pCi-yr)andconverting0tohin pC1/sec yields 5, = 31.54 .E h DFG (3-18) 3 gg, i rurem) Ci-sec) pC1) arem) yr pCi-yr sec C1 3-24

Eq. 3-18 is rewritten in the form: co i

DFGje,

( 3-19) i - mrem) uC1) arem-sec) yr sec pCi-yr where DFG y, = 31.54 DFG (3-20) ko Iarem-sec) (Ci-sec) Imrem) pCi-yr-pci-yr Ci Because maximum annual average meteorology is used (see Section 3.10), and a number of biological pathways are assumed then, Dco, has physical meaning only if Q is averaged over a year, otherwise D, only has regulatory 3 c meaning. The selection of critical receptor, outlined in Section 3.10 is inherent in Method I, as are the maximum expected off-site atmospheric dispersion factors. Should Method II be needed, the analysis for critical receptor critical pathway (s) and annual average atmospheric dispersion factors may be performed with actual meteorologic and latest census data. Because of the choice of atmospheric dispersion factors and pathways, it is expected that Method I results always will exceed Method II calculations. Either method provides adequate margin to ensure that the annual average concentrations based on organ dose from 10CFR20.106(d) are not exceeded and that the derived instantaneous release monitor readings are conservative. 3-25

3.7 Method to Calculate the Gamma Air Dose from Noble Gases Technical Specification 3.8.F.1 limits the gamma dose to air from noble gases at any location at or beyond the site boundary to 5 mrad in any quarter and 10 mrad in any year. Dose evaluation is required at least once per month. Use Method I first to calculate the gamma air dose for the plant vent stack releases during the period. Method I applies at all dose levels. Use Method II if a more accurate calculation is needed, or if Method I cannot be applied. 3.7.1 Method I The gamma air dose from plant vent stack releases is: 0,}r=0.023 1 Qj DF} (3-21) i (mrad) (DCi- ) (gg) (mrad-m ) Ci-m pCi-yr where: Q = total activity (Curies) released to the atmosphere via the plant vent stack of each radionuclide "i" during the period of interest. DF{r gamma dose factor to air for radionuclide "i". See Table 1.1-10 Equation 3-21 can be applied under the following conditions (otherwise justify Method I or consider Method II): 1. Normal operations (not emergency event), and 2. Noble gas releases via the plant vent stack to the atmosphere. l L 3-26 l l

3.7.2 Method II If Method I cennot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, than Methcd II may ue applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. 3.7.3 Basis for Method I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. Method I may be used to show that the Technical Specification which limits off-site gamma air dose from gaseous ef fluents (3.8.A.1) has been met for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50, Appendix I, Supsection B.1, which limits the estimated annual gamma air dose at unrestricted area locations. Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC within 30 days. l The concept of " increment in annual average dose", denoted AD was introduced in Section 3.0. For any noble gas release, in any period, the increment in annual average dose is taken f rom Equations B-4 and B-5 of j Regulatory Guide 1.109 with the added assumption that Dfinite = DDM/[UQD AD{r = 3.17E+04 [X/Q]T EQ DF{ (3-22) a g l 1 3 d (mrad) ( -{,r ) (sec/m ) (C1) ( ) c _p where: [X/Q]T = maximam annual average gamma atmospheric dispersion factor = 7.2E-07 (sec/m ) l i 3-27 L

,~ '~ 0 = number of curies of noble gas "1" released 9 which leads to: 0.023 E0 DF} (3-21) D,jr = 9 1 3 (arad) (DCi-vr) (C1) (e. rad-m ) 3 pCi-yr Ci-m The only difference between Method I and Method II is that Method II would use actual meteorology to determine [X/Q]T rather than use the maximum annual average value obtained for the years 1978 to 1982. l ~ 3-28

3.8 Method to Calculate the Beta Air Dose from Noble Gases Technical Specification 3.8.F.1 limits the beta dose to air fio.1 noble gases at any location at or beyond the site boundary to 10 mrad in any quarter and 20 mrad in any year. Dose evaluation is required at least once per 31 days. Use Method I first to calculate the beta air dose for the plant vent stack releases daring the period. Method I applies at all dose levels. Use Method II if a more accurate calculation is needed or if Method I cannot be applied.

3. ti.1 Method I The beta air dose from plant vent stack releases is:

0.02 b0 DF (3-23) 0 0 = 1r 3 i (a. rad) (o r) (C1) (mra _ ) p Ci-m where: 0 DF = eeta dose factor to air for radionuclide "1". See Table 1.1-10 Q = total act Wity (Curies) released to the atmosphere via the plant g vent stack of each radionuclide "i" durir.g the period of interest. Equation 3-23 can be applied under the following conditions (otherwise justify Method I or consider Method II): 1. Normal operations (not emergency event), and 2. Noble gas releases via the plant vent stack to the atmosphere. 3-29

3.8.2 !!ethod II If Method I cannot be applied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be applied. Method II consists of the models, input data and assumptions in Regulatory Guide i.109, Rev.1 (Refe ence A), except where site-specific models, data or assumptions are more applicable. 3.8.3 Basis for Method-I This section serves three purposes: (1) to document that Method I complies with appropriate NRC regulations. (2) to provide background and training information to Method I users, and (3) to provide an introductory user's guide to Method II. The methods to calcuhte beta air dose parallel the gamma air dose methods in Scction 3.7.3. Only the differences are presented here. Method I may be used to show that the Technical Specification which limits off-site beta air dose from gaseous ef fluents (3.8.A.1) has been u.et for releases over appropriate periods. This Technical Specification is based on the Objective in 10CFR50, Appendix I, Subsection B.1, which limits the estimated annual beta air dose at unrestricted area locations. Exceeding the Objective does not immediately limit plant operation bLt requires a report to the NRC within 30 days. For any noble gas release, in any period, the increment in annual average dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109: 0 0 AD = 3.1 JE+04 X/Q E Qj DF (3-24) air i 3 (mrad) (DCi-vr) (sec) (ci) (mead-m ) ci-sec ,3 pci-yr i 3-30

substituting X/Q = Maximum annual average undepleted atmospheric dispersion factor. = 6.3E-07 sec/m 4 W* have 0.02 $Q DF (3-23) D = 3 jp 1 3 (mrad) (cCi-r) (C1) ("C _yrad-m ) Ci-m 3-31

3.9 Method to Calculate the Critical Oraan Dose from Iodines. Tritium and Particulates Technical Specification 3.8.G.1 limits the critical organ dcte to a Member of the Public from radioactive Iodines, Tritium, and particulates with half-lives greater than 8 days (hereafter called "I+P") in gaseous effluents to 7.5 mrem per quarter and 15 mrem per year. Technical Specification 3.8.M.1 limits the total body and organ dose to any real member of the public from all station sources (including gaseous effluents) to 25 mrem in a year except for the thyroid, which is limited to 75 mrem in a year. Use Method I first to calculate the critical organ dose from a vent stack release as it is simpler to execute and more conservative than Method II. Method I is conservative for total body, critical organ, and thyroid dose greater than 0.1 mrem. Use Method II if a more accurate calculation of critical organ dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied, or if the majority of the release is Iodine and the 75 mrem limit for Specification 3.8.L.1 needs to be evaluated. 3.9.1 Method I }[ 0 DFG ( ~ '} D, = 9 ico c i (mrem) (C1)(*[{") Q = Total activity (Ci) released to the atmosphere of radionuclide g "1" during the period of interest. For strontiums, use the most recent measurement. DFG = Site-specific critical organ dose factor (mrem /C1). For each y radionuclide it is the age group and organ with the largest dose facter. See Table 1.1-12. 3-32

Equation 3-25 can be applied under the following conditions (otherwise, justify Method I or consider Method II): 1. Normal operations (not emergency event), 2. I+P releases via the plant stack to the atmosphere, and Any continuous or batch release over any time period. 4. 3.9.2 METHOD II If Method I cannot be applied, or if the Method 7 dose exceeds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of the models, input data and assumptions in Regulatory Guide 1.109 Rev.1 (Reference A), except where site-specific models, data or assumptions are more applicable. The base case analysis, documented below, is a good example of the use of Method II. It is an acceptable s!arting point for a Method II analysis. 3.9s3 Basis for Method I This section serves three purpotes: (1) tc document that Method I complies with appropriate NRC regulations, (2) to provide background and training information to Method I users, and (3) to provide ar. introductory user's guide io Method II. Method I may be used to show that the Technical Specifications which limit off-site organ dose from gases (3.8.G.1 and 3.8.L.1) have been met for releases over the appropriate periods. These Technical Specifications are based on Objectives and Standards in 10CFR and 40CFR. Technical Specification 3.8.G.1 is based on the ALARA Objectives in 10CFR50, Appendix I, Subsection II C. Technical Specification 3.8.M.1 is based on Environmental Standards for Uranium Fuel Cycle in 40CTR190 (hereaf ter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents. These methods apply only to I+P in gaseous effluents contribution. 3-33

Exceeding the Objective or the Standard does not immediately limit plant operation but requires a report to the NRC within 30 days. In addition, a waiver may be required. Method I was developed scch that "the actual exposure of an individual... is unlikely to be substantially underestimated" (10CFR50, Appendix I). The use below of a single " critical receptor" provides part of the conservative margin to the calculation of critical organ dose in Method I. Method II allows that actual individuals, with real behaviors, be taken into account for any given release. In fact, Method I was based on a Method II analysis of the critical receptor for the annual average conditions. For purposes of complying with the Technical Specifications 3.8.G.2 maximum annual average atmospheric dispersion factors are appropriate for batch and continuous releases. That analysis was called the " base case"; it was then reduced to form Method I. The base case, the method of reduction, and the assumptions and data used are presented below. The steps performed in the Method I derivation follow. First, in the bas? case, the dose impact to the critical receptor in the form of dose factors (mrem /C1) of 1 curie release of each I+P radionuclide to gaseous - . effluents was derived. Then Method I was determined using simplifyir.g and further conservative assumptions. The base case analysis uses the methods, data and assumptions in Regulatory Guide 1.109 (Equations C-2, C-4 and C-13 in Reference A). Tables 3.9-1 and 3.9-2 outline human consumption and environmental parameters used in the analysis. It is conservatively assumed that the critical receptor lives at the "maximun of f-site atmospheric dispersion factor location" as defined in Section 3.10. However, he is exposed, conservatively, to all pathways (see Section 3.10). The resulting site-specific dose factors are for the maximum organ and the age group with the highest dose factor for that organ. These critical organ, critical age dose factors are given in' Table 1.1-12. For any gas release, during any period, the increment in annual average dose-from radionuclide "1" is: ADj,= Q DFG,, (3-26) g 3, _ _.

where DFGgg, is the critical dose factor for radionuclide "1" and Qg is the activity of radionuclide "i" released in curies. Method I is more conservative than Method II in the regica of the Technical Specification limits because it is based on the fallowing reduction of the base case. The dose factors DFG used in Method I were chosen from g the base case to be the highest of the set for that radicr.uclide. In effect each radionuclide is conservatively represented by its own critical age group and critical organ. Because of the assumptions about receptors, environment, and radionuclides; and because of the low Objective and Standard, the lack of immediate restriction on plant operation, and the adherence to 10CFR20 concentrations (which limit public health consequences) a failure of Method I (i.e., the exposure of a real individual being underestimated) is improbable and the conseq;ences of a failure are mininal. 3-35

i Table 3.9-1 Environmental Parameters for Gaseous Effluents at Vermont Yankee (Derived from Reference A) Vegetables Cow Milk Goat Milk Meat Variable Stored ~-Leafy Pastu_re Stored Ea1ture Stored Pasture Stored 2 YV Agricultural (Kg/M ) 2. 2. 0.75 2. 0.75 2. 0.75 2. Productivity 2 P Soil Surface (KG/M ) 240. 240. 240. 240. 240. 240. 240. 240. Density T Transport Time (HRS) 48. 48. 48. 48. 480. 480. to User T8 Soil Exposure (HRS) 131400. 131400. 131400. 131400. 131400. 131400. 131400. 131400. Time TF Crop Exposure (HRS) 1440. 1440. 720. 720. 720. 720. 720. 720. Time to Plume 7H Holdes After (HRS) 1440. 24. O. 2160. O. 2160. O. 2160. Harvest QF Animals Daily (KG/ DAY) 50. 50. 6. 6. 50. 50. Feed FP Fraction of Year 0.50 0.50 0.50 on Pasture FS Fraction Pasture 1. 1. 1. when on Pasture FG Fraction of Stored 0.76 Veg. Grown in Garden FL Fraction of Leafy 0.50 Veg. Grown in Garden FI Fraction Elementil Iodine = 0.5 3 H Absolute (gm/M ) Humidity = 8.00 3-36

Table 3.9-2 Usage Factors for Various Gaseous Pathways at Vermont Yankee (from Regulatory Guide 1.109 Table E-5) Age Leafy Group Veaetables Veaetables. Milk Meht Jnhalation (kg/yr) (kg/yr) (1/yr) ( kg/y.-) (m /yr) Adult 520.00 64.00 310.00 110.00 8000.00 Teen 630.00 42.00 400.00 65.00 8000.00 Child 520.00 26.00 330.00 41.00 3700.00 Infant 0.00 0.00 330.00 0.00 1400.00 l l 3 -37

3.10 Receptor Points and Annual Average Atmospheric Dispersion Factors for ImDortant Exposure Pathways The gaseous effluent dose methods have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else. The following exposure pathways to gaseous effluents listed in Regulatory Guide 1.109 (Reference A) have been considered: 1. Direct exposure to contaminated air; 2. Direct exposure to contaminated ground; 3. Inhclation of air; 4. Ingestion of vegetables; 5. Ingestion of gott's milk; and 6. Ingestion of meat. Section 3.10.1 details the selection of important off-site locations and receptors. Section 3.10.2 describes the atmospheric model used to convert meteorologic data into atmospheric dispersion factors. Section 3.10.3 presents the maximum atmospheric disper: ion factors calculated at each of the off-site receptor locations. l 3.10.1 Receptor Locations Three important receptor locations are considered in the dose and dose ( rate equations for gaseous radioactive effluents. They are: l 1. The point of maximum gamma exposure; l 2. The point of maximum ground level air concentration of l radionuclides; and I 3-38

3. The point of maximum ground level air concentration of radionuc? ides wnere a real milk animal exists. The point of maximum gamma exposure (S sector, 400 meters) was determined by finding the maximum annual average gamma X/Q at any of f-site location. The location of the maximum ground level air concentration of radionuclides (WNW sector, 2415 meters) was determined by finding the maximum annual average undepleted X/0 at any off-site location. The point of maximum ground level sir concentration of radionuclides where a real milk animal exists (WNW sector, 30r0 meters) was determined by finding the maximum depleted X/Q at a real milk animal location. 3.10.2 Vermont Yankee Atmospheric Dispersion Model The annual average atmospheric dispersion factors are computed for routine (long-term) releases using Yankee Atomic Electric Company's (YAEC) AE0LUS Computer Code (Reference B). AE0LUS is based, in part, on the straight-lir.e airflow model discussed in Regulatory Guide 1.111 (Reference C). The valley in which the Ilant is located is considered by the model. AEOLUS produces the following annual average atmospheric dispersion factors for each location: 1. Undepleted X/Q dispersion factor-for evaluating ground level concentrations; 2. Depleted X/Q dispersion factors for evaluating ground level concentrations; 3. Gamma X/Q dispersion factors for evaluating gamma dose rates from a sector averaged finite cloud (multiple energy undepleted source); and 4. D/Q deposition factors for evaluating dry deposition of elemental radioiodines and other particulates. 3-39

Gamma dose rate is calculated throughout this ODCM using the finite cloud model presented in " Meteorology and Atomic Energy - 1968" (Reference E,_ Section 7-5.2.5. That model is implemented through the definition of an T effective gamma itmospheric dispersion factor, [X/Q ) (Reference B, Section 6), and the replacement of X/Q in infinite cloud dose equations by the T [X/Q ). . 3.10.3 Annual Average Atmospheric DisDersion Factors for Receptors Actual measured meteorological data for the five-year period,1978 through 1982, were analyzed to determine all the values and locations of the maximum off-site annual average atmospheric dispersion factors. Each dose and dose rate calculation incorporates the maximum applicable off-site annual average atmospheric dispersion factor. The values used and their locations are summarized in Table 3.10-1. Table 3.10-1 also indicates which atmospheric ~ dispersion factors are used to calculate the varicus doses or dose rates of interest. l l L l t l L-I 3-40

Table 3.10-1 Vermont -Yankee Dilution Factors Dose to Critical Dose Rate to Individual 09:e to Air Orean Total Body Skin Critical Organ Gamma Beta Thyroid r X/Q d:pleted (5) 4.6E-07(3) 4.6E-07(3) m T X/Qundepleted(SIC) 6.3E-07(2) 6.3E-07(2) i m" D/Q (1 ) 8.9E-10(3) 8.9E-10(3) 7 R III III 7.2E-07 II) X/QT (5) 7.2E-07 7.2E-07 3 a i I 4 (1) Maximum ganum exposure point: S sector, 400 meters (0.25 miles). (2) Maximum ground level concentration: WNW sector, 2415 meters (1.50 miles). (3) Worst real iallk animal concentration: WNW sector, 3000 meters (1.86 miles). 3-41 i

3.11 Method to Calculate Direct Dose from _ Plant Operation Technical Specification 3.8.M.1 restricts the dose to the whole body or any organ to any member of the public from all station sources (including direct radiation from the turbine and reactor building) to 25 mrem in a calendar year (except the thyroid, which is limited to 75 mrem). 3.11.1 Method The maximum contribution of direct dose to the whole body or to any organ is: 1.29E-06 E D = d (3-27) (mrem) (" ) (MW,h) e whore: gross electric output over the period of interest E (MW h) e 1.29E-06 = (2.14E-06 mR/MW h) (0.6 U. rem /mR) e 3.11.2 Basis of Method The major source of direct radiation (including sky shine) from the station is due to N-16 decay in the turbine building. Because of the orientation of the turbine building on the site, and the shielding effects of l the adjacent reactor building, only the seven westerly sectors (SSW to NNW) see any significant direct radiation. i High pressure ionization chamber (HPIC) measurements have been made in the plant area in order to estimate the direct radiation from the station. The chamber was located at a point along the west site boundary which has been determined to receive the maximum direct radiation from the plant. Using l ( measurements of dose rate made while the plant operated at different power levels, f rom shutdown to 100 percent, a gross electric output to dose rate 3-42

conversion factor of 2.14E-06 mR/MWh has been derived at the location of the nearest resident based on measurements at the site boundary. Field measurements of exposure, in units of Roentgen, were modified by multiplytag by 0.6 to obtain whole body dose equivalents in units of rem in accordance with recommendations of HASL Report 305 (Reference F) for radiation fields resulting from N-16 photons. Therefore, knowing the gross megawatts generated during the period of interest one may obtain a conservative estimate of the maximum dose from direct radiation at the nearest resident throegh the use of Equation 3-27. 3.12 Cumulative Doses Cumulative Doses for a calendar quarter and a calendar year must be maintained to meet Technical Specifications 3.S.8.1, 3.8.F.1 and 3.8.G.l. In addition, if the requirements of Technical Specification 3.8.M.2 dictate, cumulative doses over a calendar year must be determined for Technical Specification 3.8.M.1. To ensure the limits are not exceeded, a running total must be kept for each release. 1 4 3-43

4.0 ENVIRONMENTAL MONITORING PROGRAM The radiological environmental monitoring stations are listed in Table 4.1. The locations of the stations with respect to the Vermont Yankee plant are shown on the maps in Figures 4-1 to 4-6. All routine radiological analyses for environmental samples are performed at the Yankee Environmental Laboratory. The Laboratory participates in the U.S. Environmental Protection Agency's Environmcntal Radioactivity Laboratory Intercomparison Studies Program for all the species and matrices routinely analysed. i i i 4-1 l

Table 4.1 Radiological Environmental Monitoring Stations

• 4 Distance From Exposure Pathway Sample loca'. ion the Plant Direction From and/or Sample and Designated Code **

Stack (km) the Plant 1. AIRBORNE (Radioiodine and Particulate) AP/CF-11 River Station 1.9 SSE No. 3.3 AP/CF-12 N. Hinsdale, NH 3.6 NNW AP/CF-13 Hinsdale Substation 3.1 E AP/CF-14 Northfield, MA 11.3 SSE AP/CF-21 Spofford Lake 16.1 NNE 2. WATERBORNE a. Surface WR-11 River Station 1.9 Downriver No. 3.3 WR-21 Rt. 9 Bridge 12.8 Upriver b. Ground WG-11 Plant Well On-site WG-12 Vernon Nursing Well 2.0 SSE WG-21 Brattleboro Country 12.1 NNW Club c. Sediment SE-11 Shoreline Downriver 0.8 On-site From SE-12 North Storm *** 0.15 On-site Shoreline Drain Outfall 3. INGESTION a. Milk TM-11 Miller Farm 0.8 WNW l TM-12 Whitaker Farm 2.6 S l TM-13 Newton Farm 5.1 SSE TM-21 Moore Farm 15.9 N b. Mixed TG-11 River Station 1.9 SSE Grasses No. 3.3 TG-12 N. Hinsdale, NH 3.6 NNE TG-13 Hinsdale Substation 3.1 E TG-14 Northfield, MA 11.3 SSE TG-21 Spofford Lake 16.1 NNE I c. Silage TC-11 Miller Farm 0.8 WNW TC-12 Whitaker Farm 2.6 S TC-13 Newton Farm 5.1 SSE TC-21 Moore Farm 15.9 N d. Fish FH-11 Vernon Pond On-site FH-21 Rt. 9 Bridge 12.8 Upriver i I 4-2

Table 4.1 (continued) Radiological Environmental Monitoring Stations

• Distance From Exposure Pathway Sample Location the Plant Direction From and/or Sample and Designated Code

_ Stack (km) the Plant 4. DIRECT RADIATION DR-1 River Station 1.9 SSE Nc. 3.3 DR-2 N. Hinsdale, NH 3.6 NNW DR-3 Hir.sdale Substation 3.1 E DR-4 Northfield, MA 11.3 SSE DR-5 Spofford Lake 16.1 NNE DR-6 Vernon School 0.6 SW DR-7 Site Boundary 0.32 SSW DR-8 Site Boundary 0.45 S DR-9 Inner Ring 1.8 N DR-10 Outer Ring 4.3 N DR-ll Inner Ring 1.8 NNE DR-12 Outer Ring 3.4 NNE DR-13 Inner Ring 1.2 NE DR-14 Outer Ring 4.2 NE DR-15 Inner Ring 1.4 ENE DR-16 Outer Ring 2.9 ENE DR-17 Inner Ring 1.3 E DR-18 Outer Ring 3.1 E DR-19 Inner Ring 3.6 ESE DR-20 Outer Ring 5.5 ESE DR-21 Inner Ring 2.1 SE DR-22 Outer Ring 3.5 SE DR-23 Inner Ring 2.1 SSE DR-24 Outer Ring 4.2 SSE DR-25 Inner Ring 2.3 S DR-26 Outer Ring 4.0 S l DR-27 Inner Ring 1.2 SSW DR-28 Outer Ring 2.4 SSW DR-29 Inner Ring 0.8 SW l DR-30 Outer Ring 2.4 SW DR-31 Inrer Ring 0.8 WSW DR-32 Outer Ring 5.0 WSW DR-33 Inner Ring 0.8 W i [ DR-34 Outer Ring 4.8 W L DR-35 Inner Ring 1.2 WNW l. DR-36 Ceter Ring 4.5 WNW DR-37 Inner Ring 2.7 NW i DR-38 Outer Ring 7.4 NW DR-39 Inner Ring 2.9 NNW DR-40 Outer Ring 4.8 NNW

• Sample locations are shown on Figures 4.1 to 4.6.
  • Station 1Xs are indicator stations and Station 2Xs are control stations (for all but the Direct Radiation stations).
    • To be sampled and analyzed semi-annually.

I 4-3

O S00 METERS %g k D ,n, og / Y 's / / i FENCELINE ] ' l i N i \\ i i SE-12 'D+-11/

• )

t 's V[RNON POND \\ STACK E* s 'IC-ll s 's s i s \\ ,m 11 's Nh \\ - INTAKE s Ill-ll p Y \\ l O9

1. 0

( / g O \\, / \\ P \\

• OISCHARGE 1

VERNON ELEMENTARY SCHOOL g l ,44 SE-Il ) f i HINSDatE. N.H. 1 \\ e' 's,' c: VERNON, V.T. p s O CONNECTICUT / l VERNON DAM ggggg / / l Figure 4-1 Environmental Sampling Locations in Close Proximity to Plant 1 4-4

I N )l r,-12 AP/CF-12 OD ,./ [# HINSDALE, N.H. t,(>, i 4, r_____- gg + V l i I Tr-13 i AP/C-13 PLANT E M l i I SEE ENLAR:EMENT IN FIGURE 4-1 VERNON DAM i [ x-12 5c-12 \\ WR-11 w l2 VERfl0N, V.T. r,-11 AP/G-ll l-K-13 W 13 / 0 1 2 3 L il y POW l KILOMETERS Figure 4-2 Environemntal Sampling Locations Within Skm of Plant 4-5 i

T l t N ) N i 'n4-21 'IC-21 ~~ V l \\i SF0FFORD LAKI ni-21 Q WR-21 'IG-21 p AP/CF-] NOG 8ACK MT. rHESTERFIELD m-21/ l3 A MARLBOR0 e r a i n BkATTLE60Rr 9 Y SEE ENLAMEMENT IN F:GURE 4-2 I I g i i n a HINSDALE I l e I GUILFORD e l WINCHESTER I lVERNONe i e PLAFT I i 1 8 i VER O T j NEW HAMPSHIRE MASSACHUSETTS MASSACHUSETTS I t e NORTHr! ELD 'IG-14 AP/CF-14 I I GREENr! ELD e 0 5 10 E I J KILOMETERS l Figure 4-3 En$ ironmental Sampling Locations Grcater than Skm from Plant 4-6

I o 500 = / METERS 0,# 7 s \\ Wp N / / I FENCELINE q ' S 's i 1 DR-33 i g N 5 s vf RN0*. P0W \\ N g STACK O s l 's, s s \\ iN!ME DR-7 \\ DR-31 \\ \\ \\

1. +,

\\, \\ ~ N.O' O!SCHARGE t VERNON ELEMENTAkY SCHOOL { g \\ j\\. g ) DR-29' i, l HINSDALE, N.H. .~ / ~~ q / m E kh VE RN0*4, 7.1. J =^- x oN7 ' y Jwf co utcTIcur I 'iERh0N DA" RlyE R Gigure 4.4 TLD Locations in Close Proximity to Plant 47

t N N NEZ l / DR-40 p DR-lO \\ DR-12 DR-2 on-39 O HINSDA N.H. / .OR-14 / \\ DR_36 DR-11, DR-9 p +., t // fR-1. _,,[ --/- .tg r 1' DR-3 E NT E ) A \\On-la m triusc ", ncust 4 4 k..' DR-19 \\ l

• 1 VERNON

.T. op.2g/ f ' / DR-32[ DR-25/ \\ DR-22 DR-26 DR-24 SSW SE il v roup KILONETERS SSE { S Figure 4-5 TLD Locations h'ithin Skm of Plant 4-8

N { s / / P0FFORD LAKE } g DR-5 \\ 4,* l* NOCSACK Mr. CH,STERFIELD A MARLBORO / o BRATTLEB0 0 EMENTINFIGUk[t.5 I ~ ~~~ ~~ DR-38 / \\ $ i f itSDAL I GUILFORD e h i 9 W DR-34 E i N /DR-20 I _lERMONT g,. ,j_ - -,, 1 -,, j_g N W HAMPSHIR ,_,Y,- g [ SACHUSET MASSACHUSETTS ,/ ** g NORTHFIELO N 4 10 / %9 ' I cps I ! / N GREENFIELD e 0 5 10 g K I 3 KILOMETERS Figure 4-6 TLD Locations Greater than 5km from Plant 4-9

5.0 SETPOINT DETERMINATIONS Chapter 5 contains the basis for plant procedures that the plant operator requires to meet the setpoint requirements of the Radioactive Effluent Monitoring Systems Technical Specifications. They are Specification 3.9.A.1 for liquids and Specification 3.9.B.1 for gases. Each outlines the instrumentation channels and the basis for each setpoint. l i l l i h l l l 5-1 l

~ 5.1 Liauid Effluent Instrumentation Setpoints Technical Specification 3.9.A.1 requires that the radioactive liquid effluent instrumentation in Table 3.9.1 of the Technical Specifications have alarm setpoints in order to ensure that Specification 3.8.A.1 is not -exceeded. Specification 3.8.A.1 limits the activity concentration in liquid i effluents to the appropriate MPCs in 10CFR20 and a total noble gas MPC. 5.1.1 Liould Radwaste Discharge Monitor (17/350) The sample tank pathways shown on Figure 6-1 are monitored by the liquid radwaste discharge monitor (17/350). Periodic batch releases may be made from the waste sample tariks, detergent waste tank or floor drain sample tank. 5.1.1.1 Method to Determine the Setpoint of the Ljould Radwaste Discharae Monitor (17/350) The instrument response (in counts per second) for the limiting concentration at the point of discharge is the setpoir.t denoted R pg, and is determined as follows: ( } "setpoint " D 1 "1 n di. l l (cps) (#) (cos-ml) I Ci ml where: d (5-2) = p* = Dilution factor (as a conservative measure, DF l a DF of at least 1,000 is used) (dimensionless) l F, = Flow rate past monitor (gpm) l-F = Flow rate out of discharge canal (gpm) d l 5-2 l

DF - Minimum allowable dilution factor (dimensionless) g C Ey (5-3) = i i MPC = MPC for radionuclide "1" f rom 10CFR20, Appendix B, Table II, g Column 2 (uti/ml) C,9 = Activity concentration of radionuclide "1" in mirture at the monitor (vC1/ml) S = Detector counting efficiency from the most recent liquid j radwaste discharge monitor i.alibration curve (cps /(pCi/ml)) 5.1.1.2 .Liouid Radwaste Discharae Monitor Set 90 int Example The following alarm setpoint example is for a discharge of the floor drain sample tank. The liquid radwaste discharge monitor has a typical counting efficiency, S), of 4.9E+06 cps per 1 pCi/ml of gamma emitters which emit one photon per disintegration. The activity concentration of each radionuclide, C,j, in the floor drain sample tank is determined by analysis of a proportional grab sample obtained at the radwaste sample sink. This setpoint example is based on the following data: i i C (vC1/ml) MPCg (vC1/ml) g Cs-134 2.15E-05 9E-06 I Cs-137 7.48E-05 2E-05 Co-60 2.56E-05 3E-05 l l-i l 5-3

l E C,9 = 2.15E-05 + 7.48E-05 + 2.56E-05 1 Iml ) InG.J.) Iml ) InGj,) di uCi m1 ml = 1.22E-04 EG.! Im1 ) C DF,gn =Ey (5-3) 1 i t-uCi-sl Iml yC1) 2.15E-05 + 7.48E-05 + 2.50C-05 9E-06 2E-05 3E-05 uCi-tr.1 Iml wC1) Idi-m1) Iml pC1) uCi-al ml yC1 =7 The minimum dilution factor, DFmin' "" 3' 9' l radionuclides in this example is 7. As a conservative measure an actual f dilution factor, DF, of 1,000 is usually used. The release rate of the floor l drain sample tank may be adjusted from 0 to 50 gpm and the dilution pumps can supply up to 20,000 gpm of dilution water. With the dilution flow taken as 18,000 gpm, the release rate f rom the floor drain sample tank may be determined as follows: d l F, g = l (gpm) (gpm) (5-4) L 5-4 1

f 1 18.000 aDe 1,000 = 18 gpm Under these conditions, the setpoint of the liquid radwaste discharge monitor is: OF setpoint " 0F 1 C,9 ($_j) 3 g gQ_{ (cps) h) (rDs11) {m1 ) pCi

1. 00 4.9E+06 1.22E-04

= (#) (CDs-ml) Ig[1) 9C1 m1 = 85.400 cps In this example, the count rate alarm of the liquid radwaste dhcharge monitor should be set at 85,400 cps above background. 5.1.1. 3 Basis fcr the liquid Radwaste Discharae Monitor SetDoint j The liquid radwaste discharge monitor setpoint must ensure that Specification 3.8.A.1 is not exceeded for the appropriate in-plant pathways. The liquid radwaste discharge monitor is placed upstream of the major source of dilution flow and responds to the concentration of rcdioactivity as follows: R =E C S (5-5) mi 11 i I' (cps) {mi ) gCDs-m1) E.1 C 9C1 i i-5-5

where: R = Response of the monitor (cps) S = Detector counting efficiency for radionuclide "1" jj (cps /(yCi/ml)) C,j = Activity concentration of radionuclide "i" in mixture at the monitor (yCi/ml) The detector calibration procedure establishes a counting efficiency for a given mix of nuclides seen by the detector. Therefore, in Equation 5-5 one represents the counting efficiency may substitute S) for Sjg, where Sj determined for the current mix of nuclides. If the mix of nuclides changes significantly, a new counting efficiency should be determined for calculating the setpoint. S EC (5-6) R = j g i uCi (cps) (cos-ml) Im1 ) pCi The MPC for a given radionuclide must not be exceeded at the point of discharge. When a mixture of radionuclides is present, 10CFR20 specifies that the concentration at the point of discharge shall be limited as follows: C E gpf 51 (5-7) i i uCi-ml Imi yci) where: C = Activity concentration of radionuclide "i" in the mixture at di the point of discharge (yCi/ml) l 5-6

T MPCg = MPC for radionuclide "1" from 10CFR20, Appendix 8. Table II, Column 2 (yC1/ml) The activity concentration of radionuclide "1" at the point of discharge is related to the activity concentration of radionuclide "1" at the monitor as follows: F di C,9 [d (5-8) C = E (m1 ) (m"l ) (E U) gpm where: C = Activity concentration of radionuclide "1" in the mixture at the di point of discharge (vCi/ml) F, = Flow rate past monitor (gpm) F = F1 w rate out of discharge canal (gpm) d Substituting the right half of Equaticn 5-8 for C in Equation 5-7 and g solving for F " m yields the minimum dilution factor needed to comply with d Equation 5-7: i F C (5-3) DF I 1 M ain I m i i I uCi-m1 l Igg) Iml pC1) gpm l where: E = Flow rate out of discharge canal (gpm) d F, = Flow rate past monitor (gpm) l 5-7 l-tw1* e + ---w--

r C,9 - Activity concentration of radionuclide "i" in mixture at the monitor (pC1/ml) MPC = MPC for racionuclide "1" from 10CFN20, Appendix 8. Table II, 9 Column 2 (pC1/ml) If F /F,is less than DFain, then the tank may not be discharged until d either F r F,or both are adjusted such that: d d p DF (5-3) min. m (E) gpm usually F /F,is greater than DFg (i.e., there is more dilution than d necessary to comply with Equation 5-7). The response of the liquid radwaste discharge monitor at the setpoint is therefore: setpoint " D in (cps) (#) (*

• )

( ) 5.1.2 Service Water Discharae Monitor (17/351) The service water pathway shown on Figure 6-1 is continuously monitored l by the service water discharge monitor (17/351). The water in this lir.e is ( not radioactive under normal operating conditions. The alarm setpoint on this monitor is set at a level which is three times the background of the instrument. The service water is sampled if the monitor is out of service or if the alarm sourds. [ Under normal operating conditions, the concentration of radiciiuclides at the point of discharge fron; the service water effluent pathway will never i exceed the maximum permissible concentration in 10CFR20, Appendix B, Table II, Column 2. l l 5-8

/; 5.2 Gaseous Effluent Instrumentation Setooints Technical Spec *fication 3.9.B.1 requires that the radioactive gaseous effluent 'nstrumentation in Table 3.9.2 of the Technical Specifications have their alarm setpointi, set to insure that Technical Specifications 3.8.E.1 and 3.8.K.1 are not exceeded. Technical Specification 3.8.E.1.a limits the activity concentration in off-site gaseous effluents to well below the appropriate MPCs in 10CFR20 by liaiting the whole body and skin dose rates to areas at or beyond the site boundary. Technical Specification 3.8.K.1 limits the gross radioactivity release rate at the steam jet air ejector (SJAE) to 0.16 C1/sec. 5.2.1-Plcut Stack Noble Gas Activity Monitors (RR-100-1 A and RP-108-18) and Augmented Off-Gas System Noble Gas Activity Monitors (3127 and 3128) The plant stack and A0G noble gas activity monitors are shown on Figure 6-2. 5.2.1.1 Method to Determine the Setooint of the Plant Stack Noble Gas Activity Monitors (RR-108-1 A and RR-108-18) and the Auamented Of f-Gas Evstem Noble Gas Activity Monitors (3127 and 3 M 8) The setpoints of the plent stack and A0G system noble gas activity monitors are determined in the sare manner. The plant stack or A0G system noble gas activity monitor response in counts per minute at the limiting off-site noble gas dose rate to the total body or to the skin is the setpoint, denoted R is the lesser of: setpoint* setroint h 1,000 S R = DFB tb g c 3 _c,3 3 mrem vCi-m ) I,,pCi,, I,sec_} I oci-yr) c ICP"I I yr-pCi-sec c,3 mrem 1 and: b3

f R,g,n = 3,000 s, f (5-1 ) oF' ICP") Iarem) Icom-cm Isec) I uCi-vr ) yr pCi ,3 mrem-sec where: R = Response of the monitor at the limiting total body dose tb rate (cpm) 500 gmrem uci m ) 1,000 = yr-pCi-sec (0.7) (1E+06) (7.2E-07) 500 = Limiting total body dose rate (mrem /yr) 0.7 = Attenuation factor that accounts for the dose reduction due to shis1 ding provided by residential structures (dimensionless) 1E+06 = Number of pCi per 901 (pCi/pC1) 7.2E-07 = [X/Q]T, maximum annual average gamma atmospheric dispersion 3 factor (sec/m ) S = Appropriate (plant stack or A0G system) detector g counting efficiency from the most recent calibration (cpm /(pC1/cc)) F = Appropriate (plant stack or A0G system) flow rate (cm /sec) DFB = Composite total body dose factor (mrem-m /pCi-yr) g EQ DFB g q i (5-11) = Ehj i 5-10

r q h = Peak release rate of noble gas "i" in the mixture averaged g by the time constant of the measuring system for each noble gas listed in Tabla 1.1-10 (yC1/sec) DFB = Total body dose factor (see Table 1.1-10) (mrem-m /pci-yr) q R = Response of the monitor at the limiting skin dose rate skin (cpm) 3,000 = Limiting skin dose rate (mrem /yr) DF' = Composite skin dost. factor (n; rem-sec/pCi-yr) E h DFj j i ( 5-12) =- Ehi i DFj = Combined skin dose factor (see Table 1.1-10) (mrem-sec/pci-yr)

5. 2.1. ? Plant Stack Noble Gas Activity Monitor Setpoint Example l

The following setpoint example for the plant stack noble gas activity l monitors demonstrates the use of equations 5-9 ani 5-10 for determining setpoints. l The plant stack noble gas activity monitors, referred to as " Stack l Gas I" (RR-108-1A) and " Stack Gas II" (RR-108-1B), consist of beta sensitive l scintillation detectors, electronics, an analog ratemeter readout, and a digital scaler which counts the detector output pulses. A strip chart recorder provides a permanent record of the ratemeter output. The monitors have typical calibration factors, S, of lE+08 cpm per 1 pCi/cc of noble g gas. The nominal plant stack flow is 7.4E+07 cc/sec ((156,000 cfm x 28,300 cc/ft )/60 sec/ min). l 5-11 L

b The release rate of each noble gas is determined by analysis of a sample of of f ;2s obtained at the steam jet air ejector (SJAE). This setpoint exarale is based on the following data (see Table 1.1-10 for DFBg andDFj): i h DFB, 3 DFj g IuCi) I I I I mrem-m mrem-sec sec oCi-vr uCi-vr Xe-138 1.03E+04 8.83E-03 7.76E-03 Kr-87 4.73E+02 5.92E-03 9.59E-03 Kr-88 2.57E+02 1.47E-02 1.00E-02 Kr-85m 1.20E+02 1.17E-03 1.61E-03 Xe-135 3.70E+02 1.81E-03 2.25E-03 Xe-133 1.97E+01 2.94E-04 3.90E-04 E h DFB j g (5-11) DFB = c EQi i E h DFBq = (1.03E+04)(8.83E-03) + (4.73E+02)(5.92E-03) g i + (2.57E+02)(1.47E-02) + (1.20E+02)(1.17E-03) + (3.70E+02)(1.81E-03) + (1.97E+01)(2.94E-04) = 9.83E+01 (yci-mrem-m /sec-pCi-yr) Eh = 1.03E+04 + 4.73E+02 + 2.57E+02 g i + 1.20E+02 + 3.70E+02 + 1.97E+01 = 1,15E+04 pCi/sec 9.83E+01 DFB " 1.15E+04 c 3 = 8.52E-03 (mrem-m / Ci-yr) 5-12

Rtb = 1,000 SghDFB (5-9) g = (1,000) (1E48) (7.4E+07) (8.52E-03) = 160,000 cpm E h OFj j I DF' = (5-12) 0 1 i E h DFj = (1.03E+04)(7.76E-03) + (4.73E+02)(9.59E-03) g i + (2.57E+02)(1.00E-02) + (1.20E+02)(1.61E-03) + (3.70E+02)(2.25E-03) + (1.97E+01)(3.90E-04) = 8.81E+01 (vCi-mrem-sec/sec vCi-yr) DF' = 8.81E @ c 1.15E+04 = 7.66E-03 (mrem--sec/vCi-yr) Rskin = 3,000 SghDF' (5-10) l = (3,000) (1E+08) (7.4E+07) (7.66E-03) ( = 530,000 cpm The setpoint, R s e esser of R and R For the

setpoint, tb skin.

noble gas mixture in this example R is less than R, , indicating that tb the total body dose rate is more restrictive. Therefore, in this example the i l 5-13

y " Stack Gas I" and " Stack Gas II" noble gas acti: ty monitors thould each be set at 160,000 cpm above background. 5.2.1.3 Basis for the Plant Stack and A0G System Noble Gas Activity Monitor. Setooints The setpoints of the plant stack and A0G system noble gas activity monitors must ensure that Technical Specification 3.8.E.1.a is not exceeded. Sections 3.4 and 3.5 show that Equations 3-5 and 3-7 are acceptable methods for determining compliance with that Technical Specification. Which equation (i.e., dose to total body or skin) is more limiting depends on the noble gas mixture. Therefore, each equation must be considered separately. The derivations of Equations 5-9 and 5-10 begin with the general equation for the response R of a radiation monitor: S C,9 ( 5-13) R = gj (cpm) (CDm m ) (u ) cm where: R = Response of the instrument (cpm) gj = Detector counting efficiency for noble gas "i" (cpm /(pCi/cm )) S C = Activity concentration of noble gas "i" in the mixture at the g 3 noble gas activity monitor (pC1/cm ) The peak release rate of each noble gas, Qg (pCi/sec), is determined by analysis of a sample of off-gas obtained at the steam jet air ejector (SJAE). Cg, the activity concentration of noble gas "1" at the noble gas activity monitor, may be expressed in terms of Qg by dividing by F, the appropriate flow l rate. In the case of the plant stack noble gas activity monitors the appropriate flow rate is the plant stack flow rate and for the A0G noble gas activity monitors the appropriate flow rate is the A0G system flow rate. 5-14

f~ h 1 (5-14) C,9 = j p I i_) I901) I I MC sec c,3 c,3 sec where: h = Peak release rate of noble gas "i" in the mixture averaged by the y time constant of the measuring systen for each noble gas listed in Table 1.1-10 F = Appropriate flow rate (cm /sec) yields: Substituting the right half of Equation 5-14 into Equation 5-13 for Cg S h h (5-15) R = gj g (cpm) (C ")( ) (sec) 3 cm The detector calibration procedure establishes a counting efficiency for a given mix of nuclides seen by the detector. Therefore, in Equation 5-15 one may substitute S for Sgg, where S represents the counting efficiency g g determined for the current mix of nuc11 des. If the mix of nuclides changes significantly, a new counting efficiency should be determined for calculating the setpoint. hEh ( 5-16) S R = g g (cpm) (C

• ) (sec)

( ) m The total body dose rate due to noble gase., is determined with Equation 3-5: 0.50 E h DFB (3-5) I = g tb 1 5-15

3 gerem) (Dri-sec) (di.) gerem-m yr 3 sec pC1-yr pCi-m where: O = total body dose rate (mrem /yr) tb 3 0.50 = (0.7E+06) x (7.2E-07) (pci-sec/pci-m ) 0.7 = attenuation factor that accounts for the dose reduction due to shielding provided by residential structures (dimensionless) 1E + 06 = number of pCi per pCi (pCi/pci) 7.2E - 07 = [X/Q]T, maximum annual average gamma atmospheric dispersion factor (sec/m ) Q = peak release rate of noble gas "i" in the mixture averaged g by the time constant of the measuring system for each noble gas listed in Table 1.1-10 (pCi/sec) DFB = total body dose factor (see Table 1.1-10) 4 3 (mrem-m /pci-yr) A composite total body gamma dose factor, DFB, may be defined such that: DFB E Q EQ DFB (5-17) = g q c j i i 3 gdi) ggC,1) (mrem-m ) arem-m pC1-yr sec sec pCi-yr Solving Equation 5-23 for DFB yields: 5-16

T-E h 0F8 i 9 ( 5-11) DFB = c EQi i Technical Specification 3.8.E.1.a limits the dose rate to the total body from noble gases at any location at or beyond the site boundary to 500 mrem /yr.-Bysetting5 equal to 500 mrem /yr and substituting DFB for DFB tb c j in Equation 3-5, one may solve for [ Q at the limiting td. ole body noble gas g dose rate: Eh= 1,000 I ~IO) j DFB i C IuCi) Imrem-uci-m ) IDCi-vr ) sec yr-pCi-sec 3 mrem-m Substituting this result for hj in Equation 5-16 yields Rtb, the response of the monitor at the limiting noble gas total body dose rate: h 1.,000 S R = DFB tb g c (cpm) (mrem uCi-m ) Icom-cm ) Isee) IDCi-Vr ) yr-pCi-sec pCi 3 3 cm mrem-m The skin dose rate due to noble gases is determined with Equation 3-7: skin i DFj (3-7) i mrem) uti) mrem-sec) yr sec .yCi-yr where: Iskin = Skin dose rate (mrem /yr) l 5-17 l-

{ F h = Peak release rate of noble gas "1" in the mixture averaged by y the time constant of the measuring system for each noble gas listed in Table 1.1-10 (uci/sec) DFj = Combined skin dose factor (see Table 1.1-10) (mrem-sec/pCi-yr) A composite combined skin dose factor, DF', may be defined such that: DFj Eh Eh DFj ( 5-19) = y g i i Imrem-sec) Idi) IuCi) I I mrem-sec pCi-yr sec sec pCi-yr Solving equation 5-19 for DF' yields: 'E h DFj j I DF' = (5-12) E0 g i Technical Specification 3.8.E.1.a limits the dose rate to the skin from noble gases at any location at or beyond the site boundary to 3,000 mrem /yr. skin equalto3,000 mrem /yrandsubstitutingDFlforDFjin By setting D Equation 3-7 one may solve for E Q at the limiting skin noble gas dose rate: j 1 E h = 3,000 ( - 0) j DF' i c uCi-vr uCi) arem) mrem-sec) sec yr Substituting this result for E h in Equation 5-16 yields Rskin, the response g of the monitor at the limiting' noble gas skin dose rate: h (5- 0) Rskin = 3,0% S DF' g 3 (cpm) (aremIIcom-cm ) I'sec) I uCi-vr ) yr pCi -3 mrem-sec cm 5-18

5.2.2 Steam Jet Air Ejector (SJAE) Noble Gas Activity Monitors (17/150A and 17/1508) The steam jet air ejector noble gas activity monitors are shown on Figure 6-2. 5.2.2.1 Method to Determine the Setpoints of the Steam Jet Air Ejector Offaas Activity Monitors (17/150A and 17/1508) The SJAE noble gas activity monitor response in counts per minute at the limiting release rate is the setpoint, denoted R tpoi d, and is determined as follows: 1.6E+05 S 1 (5-21) R = tp & t g (m/hr) ("sec) (mR-cc ) (cc ) sec hr vci where: R = Response of the monitor at the limiting release rate (mR/hr) setpoint 1.6E+05 = Limiting release rate for the SJAE specified in Technical Specification 3.8.K.1 (vCi/sec) S = Detector counting efficiency from the most recent I calibration ((mR/hr)/(vC1/cc)) 7 F = SJAE gaseous discharge flow (cc/sec) 5.2.2.2 Basis for the SJAE Noble Gas Activity Monitor Setpoint l l The SJAE noble gas activity monitor setpoint must ensure that Technical Specification 3.8.K.1 is not exceeded. The derivation of Equation 5-21 is straightforward. Simply taking equation 5-16 ar.d substituting the limiting releaserateattheSJAEforhyieldsEquation5-21,thesetpointequationfor i the SJAE noble gas activity monitor. l 5-19

1 6.6 LIQUID AND GASEOUS EFFLUENT STREAMS. RADIATION MONITORS AND RADWASTE TREATMENT SYSTEMS Figure 6-1 shows the liquid ef fluent streams, radiation monitors and the appropriate Liquid Radwaste Treatment System. Figure 6-2 shows the gaseous effluent streams, radiation monitors and the appropriate Gaseous Radwaste Treatment System. 6.1 In-Plant Liauid Effluent Pathways The liquid Radwaste System collects, processes, stores and disposes of all radioactive liquid wastes. Except for the cleanup phase separator equipment, the condensate backwash receiving tank and pump and waste sample tanks, floor drain sample tank and waste surge tank, the entire Radwaste System is located in the radwaste building. The Radwaste System is controlled from a panel On the Radwaste Building Control Room. The Liquid Radwaste System consists of the following components: 1. Floor and equipment drain system for handling potentially radioactive wastes. 2. Tanks, piping, pumps, process equipment, instrumentation and auxiliaries necessary to collect, process, store and dispose of potentially radioactive wastes. i ( The liquid radwastes are classified, collected, and treated as either high purity, low purity, chemical or detergent wastes. "High" purity and " low" purity mean that the wastes have low conductivity and high conductivity, respectively. The purity designation is not a measure of the amount of radioactivity in the wastes. t High purity liquid wastes are collected in the 25,000-ga11on waste l collector tank. They originate from the following sources: 1. Drywell equipment drains 2. Reactor building equipment drains l 6-1

3. Radwaste building equipment drains 4. Turbine building equipment drains 5. Decanted liquids from cleanup phase separators 6. Decanted liquids from condensate phase separators 7. Resin rinse Low purity liquid wastes are collected in the 25,000-gallon floor drain collector tank. They originate from the following sources: 1. Dryw211 floor drains 2. Reactor building floor drains 3. Radwaste building floor drains 4. Turbine building floor drains 5. Other floor drains in RCA (e.g., A0G and Service Building, Stack, etc.) Chemical wastes are collc:ced in the 4,000-gallon chemical waste tank and then pumped to the floor drain ecliector tank. Chemical wastes arise from the chemical laboratory sinks, the laboratory drains and sample sinks. Radioactive decontamination solutions are classified as detergent waste and collected in the 1,000-ga!1on detergent waste tank. Once the wastes are collected in their respective waste tanks, they are processed in the most efficient manner and discharged or reused in the nuclear system. From the waste collector tank, the high purity wastes are processed in one of three alternative filter demineralizers and then, if needed, in one l j " polishing" demineralizer. After processing, the liquid is pumped to a waste sample tank for testing and then recycled for additional processing, transferred to the condensate storage tank for reuse in the nuclear system or discharged. I The low purity liquid wastes are normally processed through the floor drain filter demineralizer and collected in the floor drain sample tank for discharge or they are combined with high purity wastes and processed as high l ~ purity wastes. 6-2

I Chemical wastes are neutralized and combined with low purity wastes for processing as low purity wastes. Although there is only one discharge pathway from the Radwaste System to the river, there are three locations within the Radwaste System from which releases can be made. They are: the detergent waste tank (detergent wastes), the floor drain' sample tank (chemical and low purity wastes) and waste sample tank (high purity wastes). The contents of any of these tanks can be released directly to the river. The liquid wastes collected in the tanks are handled on a batch basis. The tanks are sampled from the radwaste sample sink and the contents analyzed for radioactivity and water purity. A release is allowed once it is determined that the activity in the liquid wastes will not exceed 10CFR20 limits. G-A discharge from any of the tanks is accomplished by first starting the sample pumps, opening the necessary valves, and positioning the flow contro1 he. The release rate in the discharge line is set between 0 and 50 gpm. The dilution pumps which supply 20,000 gpm of diluticn water are then started. An interlock does not allow discharge to the river when dilution ( water is unavailable. The effluent monitor (No.17/350) in the discharge line provides an l additional check during the release. The alarm or trip setpoint on the l monitor is set according to the Technical Specification Limits and an analysis of the contents of the tank. The monitor warns the operator if the activity of the liquid waste approaches regulatory limits. In response to a warning l' signal from the monitor, the operator may reduce the flow rate or stop the discharge. L 6.2 _In-Plant Geseous Effluent Pathways The gaseous radwaste system includes subsystems that dispose of gases I f rom the main condenser air ejectors, the startup vacuum pump, the gland seal condenser, the standby gas treatment system and station ventilation exhausts. 6-3

The processed gases are routed to the plant stack for dilution and elevated release to the atmosphere. The piant stack provides an elevated release point for the release of waste gases. Stack drainage is routed to the liquid radwaste collection system through loop seals. The air ejector advanced off-gas subsystem (A0G) reduces the ejector ~ radioactive gaseous release rates to the atmosphere. The A0G System consists of a hydrogen dilution and recombiner subsystem, a dual moisture removal / dryer subsystem, a single charcoal adsorber subsystem and dual vacuum pumps. Equipment is located in shielded compartments to minimize the exposure of maintenance personnel. Radioactive releases from the air ejector off-gas system consist of fission product noble gases, activation product gases, halogens and particulate daughter products from the noble gases. The particulates and halogens are effectively removed by the charcoal beds and high efficiency particulate filters in the advanced off-gas system. The activation product gases that are generated in significant quantities have very short half-lives and will decay to low levels in the holdup pipe, as well as in the adsorber beds. The noble gases, therefore, are expected to provide the only significant coritribution to off-site dose. The charcoal off-gas syste:n is designed to provide holdup of 24 hours for krypton and 16.6 days for xenon at a condenser air inleakage rate of 30 scfm. Steam dilution, pro ss control, Band instrumentation systems are l designed to prevent are'xplosive mixture of h'ydrogen from propagating beyond the air ejector stages $ An explosive mixture of }iydrogen should never exist in the recombiner subsystem, "30-mt~ ute' delay pipe, condenser / dryer, or n charcoal adsorber beds. To prevent a hydrogen' explosion in the ~ recombiner/prehe'ater and upstream lines during. shut'down, 'the residual of f-gas steammixturecontaininghydrogen'ishrgedwithsteamorcir. Starting procedures insure sufficient-steam is introduced upstream of the preheater to dilute any hydrogen entering the'adv3ficed off-gas, system as the air ejector L line is prepared for operation-To prevent operating uncafely, instrumentation is used to detect an explosive mixture.- N 6-4 l o

Hydrogen control is accomplished by providing redundant hydrogea analyzers on the outlet from the recombiner system. These analyzers initiate recombiner system shutdown and switchover if the hydrogen concentration at the system outlet exceeds 2*6 by volume. During an automatic shutdown, two me.fn air process valves close to isolate the recombiner system. Additionally, the recombiner bed temperatures and recombiner outlet temperature provide information about recombiner performance to insure that inflammable hydrogen mixtures do not go beyond the recombiner. Should a number of unlikely events occur, it would be hypothetically possible for a hydrogen exolosion to occur in the off-gas system. Such an explosion within the recombiner system could propagate into the large "30-minute" delay pipe, through the condenser / dryer subsystem, and into the charcoal adsorber tanks. However, the recombiner/adsorber subsystems, piping, and vessels are designed to withstand hydrogen detonation pressures of 500 psi at a minimum so that no loss of integrity would result. Furthermore, the seven tanks of charcoal would significantly attenuate a detonation shock wave and prevent damage to the downstream equipment. During normal operation, the dryer /adsorber subsystem may be bypassed if it becomes unavailable provided the releases are within Technical Specification limits. With the dryer /adsorber subsystem bypassed, the air ejector off-gas exhausts through the recombiner/ condenser subsystems, and the 30-minute delay pipe. The off-gas mixture combines with steam at the air ejector stage to prevent an inflammable hydrogen mixture of 4% by volume from entering the downstream hydrogen recombiners. Approximately 6400 lb/hr of steam introduced at the second stage air ejector reduces the concentration of hydrogen to less than 3's by volume. The recombiner subsystem consists of a single path leading from the l hydrogen dilution steam jet ejectors to two parallel flow paths for hydrogen recombination. Each recombination subsystem is capable of operating independently of the other and each is capable of handling the condenser ~ l off-gas at a startup design flow rate of 1600 lb/hr air and the normal off-gas 6-5

design flow rate of 370 lb/hr. The major components of each recombiner flow ' path are a preheater, a hydrogen-oxygen recombiner, and a desuperheating condenser. The preheater assures that the vapor entering the hydrogen-oxygen recombiner is heated to approximately 300 F. At this terr.perature, the water I vapor in the stream becomes superheated steam, thereby protecting the recombiner catalyst. During passages through the recombiner, the recombination of H and 2 0 in an ex thermic reaction increases the stream temperature to 2 approximately 520 F. This recombination results in a maximum effluent H 2 concentration of 0.lt, by volume. L The desuperheating condenser is designed to remove the heat of recombination and condense the stearn from the remaining off-gas. The condensers 61scharge the off-gas through moisture separators into the initial portion of an underground 24-inch diameter delay pipe which allows for 40's of the total system holdup volume. The pipe slopes away from the off-gas particulate (HEPA) filters in both directions for drainage purposes. Loop seals prevent gas escaping through drainage connections. Shorter lived radionuclides undergo a substantial decrease in activity in this section of the system. The preheaters/recombiners operate at pressures sligntly abeve at;.+0 spheric; the condenser and the subsystems that follow operate at subatmospheric pressures. Particulate (HEPA) filters with flame suppressant prefilters are located at the exit side of the delay pipe ahead of the moisture reme. val subsystem to remove racioactive particulates generated in the delay pipe. In the moisture removal / dryer subsystem, the moisture of the gas is reduced to increase the effectiveness of the charcoal adsorber beds downstream. The subsystem consists of two parallel cooling condensers and gas L dryer units. Each condenser is cooled by a mechanical glycol / water refrigeration system that cools the off-gas to -40 F as it removes bulk l

moisture. The dryer is designed to remove the remaining moisture by a l

6-6

3_ ,5 ) q . \\ .r. l \\ ,d )' aa x \\ n; molecula'r'hieve desihcant to a dew point of less 'than -40 F (1% RH). One of I l [the dryers adsorbs mois.ture from the of f-gas; the other desorbs moisture by i " ^~ irculiting heated air _ through the bed in closed cycle. c ~ jhe nixed reirr'igerant/ dryer conceit, improves the reliability of the i y system. If the refrigerant' system fails, the.two dryer beds operate in parallel to remove the moisture and maintain the off-gas near the cesign dew j[ point-(-40 F). If the drye'r beds fail, the -40 F dew point air leaving -the mechanical-system can: enter the guard bed for over 6 hours without ~ affecting the performance of the charcoal beds downstream. .= s. 1. (The t.harc~oal[adsorber subsystem consists of seven tanks of charcoal precedediby a smaller cha'rcoal guard bed upstream. The guard bed protects the seven main tanks;from excessive raofoactivity, levels or moisture in the event ^ of'.a malfunction 1 upstream in the moistu.re removal subsystem. The guard bed ~ al,so removes compounds which might hinder noble' gas delay.~ The seven tanks x:.. hold a' minimum of approximately 90,000 poundsfof charcoal. The'first two main' tanks can be bypassed and used for' storing a " batch of high activity" gas' for static decay. The remaining five are all in series wit.), nobypassingfeatures{sothattheoff-gastothestackmust'bedelayed. c Redundant particulate (HEPA) af_ter-filters are used to remove charcoal fines prior to the vatuum pumps. i 'A' water-sealed vacuum pump boosts the gas stream pressure to slightly -over-atmospheric pressure before it. is vented through the stack. lo assure 'mmaintainingconstantNeratingpres'suresinthesystem,amodulatingbypass valve.will recirculate process gas around.the pump as required. During ~ Mperiods of high ilow rates, both pumps can be operated in' parallel. Discharge of 'the' vacuum pemp then passes through the remaining 60"6 of 'the delay pipe ' prior to'being vented through ths sfation stack. The gland seal off-gas subsystem collects gases from the gland seal _ condenser' and the a r.hanical vacu'um pump and passes them through a charcoal c. ~ 'g s . 6-7 ~ 9 ..A ~* 4 ,e,. .,,y,. 4_.,y.,,, ,.,,.-.,..,.y_,,__,, p,,,

filter (if required) and then through holdup piping prior to release to the stack. The gases from the gland seal condenser system are discharged to the atmosphere via the ventilation stack after passing through the filter for iodine removal (if required) and then through the same 1-3/4 minute holdup piping that is used for the startup vacuum pump system. One automatic valve on the discharge side of each steam packing exhauster closes upon the receipt of high level radiation signal from the main steam line radiation monitoring subsystem to prevent the release of excessive radioactive material to the atmosphere. The exhausters are shut down at the same time the valves close. In addition, the mechanical vacuum pump is automatically isolated and stopped by a main steam line high radiation signal. The filter assembly is located in the air ejector room. The release of significant quantities of gaseous and particulate radioactive material is prevented by the combination of the design of the air ejector advanced off-gas system and automatic isolation of the system from the stack. Gas flow from the main condenser stops when the air ejectors are automatically isolated from the main condenser by either a high radiation signal in the main steam line or by high temperature and/or pressure signals from the A0G System. The gland seal off-gas subsystem is automatically isolated and stopped by a main steam line high radiation signal. In addition, monitoring the stack release provides a backup warning of abnormal conditions. 1 6-8 o

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REFERENCES A. Regulatory Guide 1.109, " Calculation 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. B. Hamawi, J. N., "AEOLUS - A Computer Code for Determining Hourly and Long-Term Atmospheric Dispersion of Power Plant Effluents and for Computing Statistical Distributions of Dose Intensity From Accidental Releases". Yankee Atomic Electric Company, Technical Report, YAEC-1120, January 19'17. C. 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. D. National Bureau of Standards, " Maximum Permissible Body Burdens and Maximum dermissible Concentrations of Radionuclides in Air and in Water for Occupational Exposure", Handbook 69, June 5, 1959. E. Slade, D. H., " Meteorology and Atomic Energy - 1968", USAEC, July 1958. F. Lowder, W. M., P. D. Raf t, and G. dePianque Burke, " Determination of N-16 Gamma Radiation Fields at BWR Nuclear Power Stations", Health and Safety Eaboratory, Energy Research and Development Administration, Report No. 305, May 1976. 6-11 L e --}}