ML20154L979
| ML20154L979 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 03/06/1988 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20153E477 | List: |
| References | |
| PROC-880306, NUDOCS 8809270014 | |
| Download: ML20154L979 (98) | |
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YANFEE NUCLEAR P0k'ER STATION OFF-SITE DOSE CALCULATION MANUAL YANKEE ATOMIC ELECTRIC COMPANY NUCLEAR SERVICES DIVISION 1671 k'ORCESTER ROAD TRAMISCHAM. MASSACHi'SETTS 01701 B
PREPARED BY/DATE REVIEhED BY/IATE APPPDVED Andrew D. Hodgdon 12/2/82 PORC 11/29/82 ORIGINAL William D. Billings 12/2/82 g-Q f D __ l 2/s,/ry PORc. &nQ ry.n y,,py'y REV1510s 1 Mtatth. ids 19ftf I
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REVISION 2 LrM [ M.
- yff, PORC Heeting 85-36
-- - -- 7 July 30, 1985
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- - # / j,A F PORC Meeting 86-16 kl#h d
REVISION 3 rebruary 25,1986 M//t 4t4 /
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REVISION 4 poRC Neeting c o-9 fd., w s'.s d,.... -.y i /;' f t
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Sept. 30, i 9 36[,.,,y c, /,. "
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REVISION 5 PORC !.;eeting ~66-06 f
REVISION 6 f,. / f gg B809270014 880630 PDR ADOCK 05000029 R
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REVISION RECORD Revision Date Description 0
12/1/82 Initial printing. Approved by PORC 11/29/82.
Submitted for USNRC approval 12/3/82.
1 3/30/84 Change in environmental monitoring sampling locations based on 1983 land use census.
Errors in Table 4.1 corrected. Maps revised.
2 7/30/85 Addition of Intercomparison Program description to Section 4.0.
Reviewed by PORC 7/30/85.
3 3/19/86 Addition of a PVS I-131 inspection limit to demonstrate compliance with Technical Specification 3.11.2.1.b.
4 5/21/86 Change in milk sampling location.
Samples no longer available at Station TM-11.
5 9/30/86 Change in food product sampling location based on 1986 land use census.
6 2/18/88 Change in liquid dose factors to reflect additional dose pathways. Change in gaseous dose factors to reflect five-year average meteorology. Change in gaseous dose rate factors to reflect a shielding factor of 1.0.
Deletion of food product location TF-12 (samples no longer required after 10/31/86).
Update of fence line location and several building names and locations in Figure 4-4.
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,o REVISION RECORD 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.
i12e Revision i to 111 6
iv to vii. ix 6
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4-1 to 4-10 1
4-1 2
3-24 to 3-26 3
4-3, 4-9, 4
4-3, 4-9. 4-10 5
1-1 to 1-17 6
3-1 to 3-50 6
4-5, 4-9. 4-10 6
5-1 to 5-13 6
6-2 6
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e DISCLAIMER OF RESPONSIBILITY This document was prepared by Yankee Atomic Electric Company ("Yankee").
The use of information contained in this document by anyone other than Yankee, or the Organization for which the document was prepared under contract, is not authorized and, with respect to any unauthorized use, neither Yankee nor its officers, directors, agents, or employees assume any obilgation, responsibility, or liability or make any warranty or i
representation as to the accuracy or completeness of the material contained in this document.
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ABSTRACT The YNPS 00CM (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 YNPS Plant Operation Review Committee (PORC) and approval of subsequent revisions 'Jy PORC (as per the Technical Specifications) this OOCH is suitable to show compliance where referred to by the Plant Technical Specifications.
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TABLE OF CONTENTS Pace REVISION REC 0RD..................................................
11 LIST OF EFFECTIVE PAGES..........................................
iii DISCLAIMER OF RESPONSIBILITY.....................................
iv ABSTRACT.........................................................
v LIST OF FIGURES..................................................
vili LIST OF TASLES...................................................
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1.0 INTRODUCTION
1-1 1.1 Summary of Methods, Dose Factors, Limits Constants, Variables and Definitions..................................
1-2 2.0-METHOD TO CALCULATE OFF-SITE LIQUID CONCENTRATIONS...............
2-1 2.1 Method............
2-1 2.2 Method to Determthe Radionuclide Concentration for Each Liquid Effluent Pathway...........................
2-2 2.2.1 Test Tank Pathway............
2-2 2.2.2 Steam Generator Blowdown Pathway...................
2-3 2.2.3 Secondary Coolant and Coolant Leakage Pathway......
2-3 2.2.4 Remaining Pathways.................................
2-3
2.3 Background
Information.....................................
2-4 3.0 0FF-SITE DOSE CALCULATION METH0DS................................
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 1
3.4 Method to Calculate the Total Body Dose Rate From Noble Gases................................................
3-14 I
3.5 Method to Calculate the Skin Dose Rate From Noble Gases....
3-19 3.6 Method to Calculate the Critical Organ Dose. Rate from 131-I, 3H and Particulates with T1/2 Greater Than 3-23 8 Days.....................................................
3.7 Method to Calculate the Gamma Air Dose From Noble Gases....
3-27 3.8 Method to Calculate the Beta Air Dose From Noble Gases.....
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TABLES OF CONTENTS (Continued) 3.9 Method to Calculate the Critical Organ Dose.from Tritium, Iodines and Particulates...................................
3-34 3.10 Critical Receptors and Annual Average Atmospheric Dilution Factors for Important Exposure Pathways...........
3-41 l
3.11 Method to Calculate Direct Dose Fron Plant Operation.......
3-45 l
4.0 ENVIRONMENTAL MONITORING LOCATIONS...............................
4-1 5.0 SETPOINT. DETERMINATIONS..........................................
5-1 5.1 Liquid Effluent Instrumentation Setpoints..................
5-2 5.2 Gaseous Effluent Instrumentation Setpoints.................
5-7 6.0 LIQUID AND GASEQUS EFFLUENT STREAMS RADIATION HONITORS AND RADWASTE TREATMENT SYSTEMS...................................
6-1 6.1 In-Plant Liquid Effluent Pathways..........................
6-1 6.2 In-Plant Gaseous Effluent Pathways.........................
6-3 REFERENCES.......................................................
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LIST OF TABLES Number Title Page 1.1-1 Sumary of concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at Yankee Plant 1-3 1.1-2 Dose factors Specific for Yankee Plant for Noble Gas Releases 1-7 1.1-3 Sumary of Radioactive Effluent Techni;al Specifications With Dose or Oose Rate Limits and Ir91ementating Method I Equations 1-8 1.1-4 Sumary of Constants 1-10
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1.1-5 Sumary of Variables 1-11 1.1-6 Definition of Terms 1-15 1.1-7 Dose Factors Specific for Yankee Plant for Liquid Releases 1-16 t
4 1.1-8 Dose and Dose Rate Factors Specific for Yankee Plant l
for Tritium, lodine, and Particulate Releases 1-17 i
2-1 Typical Radionuclides Released From Test Tanks 2-6 i
3.2-1 Environinental Parameters for Liquid Effluents at J
I Yankee Plant 3-9 3.2-2 Age Specific Usage Factors for Various Liquid Pathways at Rowe 3-10 3.9-1 Age Specific Usage Factors 3 38 4
3.9-2 Environmental Parameters for Gaseous Effluents at Yankee Plant 3-39 3.10-1 Yankee Nuclear Power Station Five-Year Average Atmospheric l
Dispersion Factors 3-44 l
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3.11-1 Estimate of Exposure Rate at Critical Receptor From VC f
l Shine, EVC, Made in Spring 1981 3-50
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4-1 Radiological Environmental Monitoring Stations 4 13 5.2-1 Sample Calculations of Gaseous Effluent Setpoint 5-12 i
5.2-2 Relative Fractions of Core Inventory Noble Gases After i
Shutdown 5-13 Revision 6 - 2/18/88 Approved By:4< i/
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1.0 INTRODUCTION
The ODCM (Off-Site Dose Calculation Manual) provides formal and approved methods for the calculation of off-site concentration, off-site doses and effluent monitor setpoints, and indicates the locations of environmental monitoring stations in order to comply with the Yankee Nuclear Power Station Radiological Effluent Technical Specifications (RETS) Sections 3/4.3.3.6, l
3/4.3.3.7, and 3/4.11, as well as the Radiological Environmental Monitoring l
Program (Section 3/4.12).
The 00CM forms the basis for plant procedures which document the off-site doses due to plant operation which are used to show compliance with the numerical guides for design objectives of Section II of Appendix ! to 10CFR Part 50.
The methods contained herein follow accepted NRC guidance, 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.
l All changes to the ODCM shall be reviewed and approved by the Plant l
Operation Review Comittee (PORC) in accordance with Technical Specification 6.15 prior to implementation. Changes made to the 00CM shall be submitted to the Commission for their information in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective.
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1.1 Summary of Methods. Dose Factors. Limits. Constants. Variables and Definitions This section summarizes the methods for the user.
The first time user should read Chapters 2 through 5.
The concentration and setpoint methods are I
documented in Table 1.1-1, as well as the Method I Dose equations.
Where more accurate Dose calculations are needed use the Method II for the appropriate dose as described in Sections 3.2 through 3.9 and 3.11.
The dos 6 factors used in the equations are in Table 1.1-2, 1.1-7, and 1.1-8 and the Regulatory
.l Limits are summarized in Table 1.1-3.
The constants, variables, and special definitions used in this ODCH are in Tables 1.1-4, 1.1-5, and 1.1-6.
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i Table 1.1-1 b
Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at Yankee Plant Equation l
No.
Maximum Ecuatio'i ENG C
2-1 Unrestricted Area FENG, y Total Fraction of MPC in I
1 Liquids Except Noble Gases 2-2 Unrestricted Area Concentration C
=IC"O NG of Noble Gases in Liquids i
f0DFL 3-1 Total Body Dose from Liquids Dtb(**} " K L
1 itb I
organ (*"*}"Kf0DFL 3-2 Organ Dose from Liquids D
1 3-3 Total Body Dose h(Y' }
- 7*03 I DFB l
tb 1
g Rate from Noble Gases t
3-4 Skin Dose Rate bs in(*yr )
I h DF' i
i
[
from Noble Gases g
I h DFG'co
.3-5 Organ Dos'e Rate from 1311, H3 and gco < mrem) i i
Particulates with T1/2 > 8 days yr g
i 1
tr(mrad)=0.25IQ,DF{
3-6 Gamma Air Dose from D
Noble Gases i
3-7 Beta Air Dose frcm D r(mrad) 0.76 I Q, DF Noble Gases i
3-6.1 Gamma Air Dose from Dgrd(mrad) = (1.23E-04)(Q,,j)3 1
y Ground Level Noble Gas Equivalent)
Releases l
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Table 1.1-1 (continued)
Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at Yenkee Plant Equation No.
Maximum Ecuati C, t
I 3-8 Organ Dose from 1311, 3H and Q OFGico D,(am m) =
y Particulates with T1/2 > 8 days e
3-9 Direct Dose Dd. (0.057 + E ) T, 0.00087 r
f 3 HPC S 5-1 Liquid ' dease Rate Reading R=f1f2 e g 5({ffsg)(500) 60 9
5-3 Gaseous Release Rate Reading Rtb "
F7.93{ffG OSF for Total Body Dose Limit g
S ({ (O s )(3000) 60 g
g
[
5-4
. Gaseous Release Rate Reading Rsk "
NG for Skin Dose Limit F
f OFj Note 1:
C
- Concentration radionuclide 1 in mixture (pC1/ml).
g e
E
. Exposure rate at critical receptor from non-VC sources as
[
I measured or estimated for the period (pR/hr).
i F
. Primary vent Stack Flow Rate (cc/ min).
{
ff.Fractionactivityofradionuclideitototalnoblegas I
activity.
l CfNG = Concentration of radionuclide "l", except noble gases, at point l
of discharge.
t Cf0. Concentration of radionuclide "i", except le gases, at point l
of discharge.
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Table 1.1-1 (continued)
Summary of Concentration and Setooint Methods, and Method I Dose Equations for Normal Operations at Yankee Plant DFj Skin dose factor for radionuclide "1".
DF{=Gammadosefactortoairforradionuclide"1",
DFf Beta dose factor to air for radionuclide "1".
DFBg Total body dose factor for radionuclide "1",
DFGgen - Site-specific, critical organ dose factor for a gaseous l
release of nuclide "i".
DFG'CO - Site-specific critical organ dose rate factor for a I
gaseous release of nucilde "1",
DFlitb - Site-specific total body dose factor for a liquid i
release of nuclide "1".
DFL
= Site-specific, traximum organ dose factor for a 11guld release cf nuclide "i".
l ISO f)
- Flow rate past test tank monitor (gem).
Flow rate past steam generator blowdown monitor (gpm).
f 2
f
- Flow rate at point of discharge (gpm).
3 K
= Deerfield River flow rate correction factor.
I HPC Composite HPC for the mix of radionuclides (pC1/ml).
g
{C g
(Eq. 5-2)
- IC I i MPC g Qg Total release (curies) for radionuclide "l".
Total release of noble gases expressed as O =-133 equivalent Xe-133 equivalent.
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(continued) p' Summary of Concentration and Setpoint Methods, and Method I Dose Equations for Normal Operations at Yankee Plant c'
L Qg Release rate (uCl/sec) for radionuclide "1".
t e
S Gaseous Instrument response factor (cpm /(pC1/cc)).
g l
S.
- Liquid instrument response factor (cpm /(pC1/cc)).
g s
Ratio of response from equal activities of radionuclide i to a g
reference radionuclide, i.e., Xe-133.
T,
. Length of exposure period (hours).
i j
t t
4 h
t i
}
l l
L t
L I
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Table 1.1-2 Dose Factors Specific for Yankee Plant for Noble Gas Releases Gamma Total Body Beta Skin Combined Skin Beta Air Gamma A'-
Dose Factor kseFactog Dose Factor DoseFact5 0 5' '*ct;
*-*3 1 (*uCi-yr ) DF0 (mrad-m ) DFY (*# ' C
'*~5'C i (*nCi-yr )
DFS (*DCl-yr )
DF' DFB i
DCi-yr i
DCi-y-i Radionuclide Ar-41 8.84E-03*
2.69E-03 1.45E-01 3.28E-03 9.30E-M 1.68E-04 2.88E-04 1.93E Kr-83m 7.56E-08 Kr-85m 1.17E-03 1.46E-03 4.56E-02 1.97E-03 1.23E-::
Kr-85 1.61E-05 1.34E-03 3.22E-02 1.95E-03 1.72r-;;
Kr-87 5.92E-03 9.73E-03 2.86E-01 1.03E-02 6.17E-;;
Kr-88 1.47E-02 2.37E-03 1.89E-01 2.93E-03 1.52E-:'
Kr-89 1.66E-02 1.01E-02 3.92E-01 1.06E-02 1.73E-C; Kr-90 1.56E-02 7.29E-03 3.16I-01 7.83E-03 1.63E-C Xe-131m 9.15E-05 4.76E-04 1.27E-02 1.11E-03 1.56E-e-Xe-133.
2.$1E-04 9.94E-04 2.66E-02 1.48E-03 3.27E-:-
Xe-133 2.94E-04 3.06E-04 1.04E-02 1.05E-03 3.53E-Et Xe-135m 3.12E-03 7.11E-04 4.62E-02 7.39E-04 3.36E-0;-
Xe-135 1.81E-03 1.86E-03 6.11E-02 2.46E-03 1.92E
c Xe-137 1.42E-03 1.22E-02 3.05E-01 1.27E-02
- 1. 51 E-0.5 Xe-138 8.83E-03 4.13E-03 1.79E-01 4.75E-03 9.21E-O
+
- 8.84E-03 = 8.84 x 10-3 l
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Table 1.1-3 Summary of Radiological Effluent Technical Specifications and Implementing Equations Technical Specification Category Method
- Limit 3.3.3.6 Liquid Effluent Alarm / Trip Setpoint Eq. 5-1 T.S. 3.11.1.1 Monitor Setpoint 3.3.3.7 Gaseous Effluent Alarm Setpoint for Eq. 5-3 T.S. 3.11.2.la Monitor Setpoint Total Body Dose Rate (Total Body)
Alarm Setpoint for Eq. 5-4 T.S. 3.11.2.la Skin Dose Rate (Skin) 3 11.1.1 Concentration Total Fraction of Eq. 2-1
< 1.0 (Liquids)
MPC Excluding Noble Gases Total Noble Gas Eq. 2-2 1 2x10-4 pC1/cc Concentration 3.11.1.2 Dose (Liquids)
Total Body Dose Eq. 3-1 1 1.5 mrem in a otr.
1 3.0 meem in a yr.
Organ Dose Eq. 3-2 1 5 mr'm in a qtr.
i 10 mrem in a yr.
3.11.1.3 Liquid Radwaste Total Body Dose Eq. 3-1 1 0.06 mrem in a me.
Treatment Organ Dose Eq. 3-2 1 0.2 mrem in a me.
3.11.2.1 Gaseous Effluents Total Body Dose Rate Eq. 3-3 1 500 mrem /yr.
Dose Rate from Noble Gases Skin Dose Rate from Eq. 3-4 1 3000 mrem /yr.
Noble Gases Organ Dose Rate from Eq. 3-5 1 1500 mremlyr.
1131, H3, and Particulates with T)/2 >8 Days j
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Table 1.1-3 (continued)
Summary of Radiological Effluent Technical Specifications r
and Implementing Equations Technical Specification Category Method' Limit 3.11.2.2 Gaseous Effluent Gamma Air Dose from Eq. 3-6 1 5 mrad in a atr.
Dose, Noble Gases Noble Gases 1 10 mrad in a yr.
1 0 mrad in a atr.
1 Beta Air Dose from Eq. 3-7 Noble Gases 1 20 mrad in a yr.
3.11.2.3 Gaseous Effluent Organ *ose from Eq. 3-8 1 7.5 mrem in a otr.
11,1. H3, and Dose, Iodir*-131 7
Tritium anc, Particulates with 1 15 mrem in a yr.
Radionuclides T1/2
>8. Days 3.11.2.4 Gaseous Radwaste Gama Air Dose from Eq. 3-6 5 0.2 mrad in a me.
Treatment Noble Gases l
Beta Air Oose from Eq. 3-7 5 0.4 mrad in a me.
Noble Gases Organ Dose from 1131. Eq. 3-8 10.3 mrem in a mc.
H3, and Particulates with Tl/2 >S days 3.11.4 Total Dose from Total Body Dose Eq. 3-1+
1 25 mrem in a yr.
All Sources Organ Dose Eq. 3-6+
Thyroid Dose Eq. 3-9 r
Eq. 3-2+
5 25 mrem in a yr.
Eq. 3-8+
Eq. 3-9 Eq. 3-2+
1 75 mrem in a yr.
Eq. 3-8+
Eq. 3-3
- More accurate meti.ods may be available (see subsequent chapters).
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c-Table 1.1-4 Survnary of Constants Constant Definition Units 0.00087
= Conversion factor mrem PR 3
4([ h) (X/Q)Y (sec/m )
0.25 3.17 x 10
. (3.17 x 10 )(7.83 x 10-0)
DCl-yr 4
3 Ci-m 3
4(h h) (X/Q) (sec/m )
0.76 3.17 x 10 DCi-
= 3.17 x 10' (2.39 1 10-5)
Cl-m{r 1.11 Average ratio of tissue to air energy absorption ratio coefficient 7.83
- 106 (pC1/uCl) + 1.0
- 7.83 x 10~0 (sec/rt )
pCl-sge 3
I UC i -m' 1
8.69
. 1.11 57 (X/Q)Y (sec/m ) 1 x 100 (pC1/ C1) 3
= 1.11(1.0)(7.83 x 10-6)< g, 39 )
pCi-sic 6
3 Cl-m l
1 x 106 (X/Q) 23.9 (1 x 10 )(2.39 x 10-5) 6 60.
Conve ston factor see min 500.
. Total body annual dose limit from ICRP2 mrem 3000.
Skin annual dose limit from ICRF2 mrem 4
C 3.17 x 10 Number of picoeuries per curie divided by number of seconds per year 6
1 x 10 Number of picocuries per microcurie Revision 6 - 2/18/88 ApprovedBy:/M
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Table 1.1-5 Summary of Variables Variable Definition Units t
C
= Total activity of all dissolved and entrained b
NG CC noble gases from all station sources ENG C
= Concentration of radionuclide "1", except CC I
noble gases. At point of discharge NO Concentration of radionuclide "t", except C
CC I
noble gases, at point of discharge 3
Concentration of radionuclide "1" uC1/m C
j or pC1/cc Dfir Beta dose to air mrad Gamx.a dose to air mrad Dair l
' Gamma dose to air from a ground level release mrad l
D
=
grd Dose to the critical organ mrem O,
a g
Direct dose mrem D
d f
Dose to the maximum organ mrem D
organ Dose to skin from beta and gamxa mrem D
skin l
Dose to the total body mrem D
=
tb l
Total body gamma 1st factor for nuclide "1"
((
DFB g
{, 3 Beta skin dose factor for nuclide "1" DFS t
g C
Combined site-specific skin dose factor
[
DF{
=
//
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Table 1.1-5 (continued)
Summary of Variables Variable Definition Units
"[*d DFj
- Gamma air dose factor for nuclide "1" 0
0F
- Beta Air dose factor for nuclide "1" 0FG
= Critical organ gaseous dose factor for mrom ICO nuclide "1" Ci 0FG'CO
- Critical organ gaseous dose rate factor mrem-sec I
for nuclide "1" pCi-yr DFL
= Maximum organ liquid dose factor for mrem ISO nuclide "1" Ci 0Fl
- Total body 11guld dose factor for erem t
itb nuclide "1" C1 l
~
b
. Critical organ dose rate due to Iodines,-
l YI CO tritium, and particulates t
b
. Skin dose rate due to noble gases
- ',[ "
t 3ggn b
Total body dose rate due to noble gases
- f*
tb D/Q
- Deposition factor for dry deposition of sec elemental radiolodines and other particulates
,2 i
Exposure rate at critical receptor from yR l
r non-VC sources as measured or estimated for hr the period k
- Limiting exposure rate at the critical (R
VC receptor from the Vapor Container during ir normal operations F
. Primary vent stack flow rate h
t i
ff
. Fractl6n activity of radionuclide i to total noble gas activity
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Table 1.1-5 (continued) t Summary of Variables Variable Definition Units F)
. Total fraction of MPC in liquid pathways ENG F
= Total fraction of MPC in liquid pathways I
(excluding noble gases) f)
. Flow rate past test tank monitor gpm
'f
. Flow rate past steam generator monitor gpm 2
f
. Flow rate at point of discharge gpm 3
N MPC
= Composite MPC for the mlx of radionuclides t
CC (see Equation 5-2)
.l C
. Maximum permissible concentration radionuclide I
CC "i" (10CFR20, Appendtx B Table 2, Column 2) d
. Total release of all' noble gases Curles Qg
. Release for radionuclide 1 Curies
[
Q
. Total release ratt, of all noble gases UCurles/sec Release rate for radionuclide "1" UCurles/sec Q
g Average undepleted dispersion factor X/Q
=
m (X/Q)0 Average depletted dispersion factor m
I Effective average gamma dispersion factor y
(X/Q)Y m
Shielding factor Ratio 5
7 Gaseous monitor response factor 5
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b Table l'.1-5 (continued)
Sumary of Variables Variable Definition Units s
Ratio of response from equal activities of g
radionuclide i to a reference radionuclide (such as Xe-123)
S Liquid monitor response factor.
g pCi T,
= Exposure period hours 40
= Conservative increment in annual average dose mrem 4
t t
I I
L l
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Table 1.1-6 Definition of Terms Table Deleted.
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e TABLE 1.1-7 Dose Factors Specific for_,Y,ankee Plant Liould leases Total Body Maximum Organ Dose Factor Dose Factor itb (mrem)
DFL,n (erem' OFL Ci g
Ci Radionuclide H-3 5.99E-04 5.99E-04 C-14 1.64E+00 8.18E+00 Cr-51 7.20E-05 1.07E-02 Mn-54 6.07E-02 5.47E-01 i
Fe-55 3.46E-02 2.11E-01 Fe-59 1.00E-01 4.53E-01 1
Co-58 4.76E-02 1.81E-01
'l Co-60 2.79E-01 9.04E-01 Zn-65 1.65E+00 2.71E+00 5r-85 2.30E-01 8.04E+00 l
Sr-90 6.97E+01 2.75E+02 Zr-95,Nb-95 1.40E-03 2.87E-01 j
i Ru-103 2.48E-03 3.57E-01 i
Ag-110m 2.32E-02 2.21E+00 Sb-124 2.62E-02 6.48E-01 I-131 8.57E-03 4.96E+00 l
I-133 6.52E-04 3.18E-01 Cs-134 1.79E+01 2.40E+01 Cs-136 2.28E+00 3.20E+00 Cs-137 1.07E+01 2.07E+01 Ba-140/La-140 3.40E-03 5.80E-02 Ce-141 7.73E-05 1.06E-01 Ce-144 1.41E-03 2.58E+00 li 5
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TABLE 1.1-8 Dose and Dose Rate Factors Specific for Yankee Plant for Iodines. Tritium an D articulate Releases Critical Organ Critical Organ Dose Factor Dose Rate Factor
-C DFG'co ("y"r*-9Ci )
ico ('Cl )
DFG i
Radionuclide H-3 7.21E-03 2.27E-01 C-14 4.38E+00 1.38E+02 Cr-51 3.44E-02 1.19E+00 1
Mn-54 3.78E+00 1.49E+02 Fe-59 3.83E+00 1.27E+02 Co-58 1.98E+00 7.06E+01 Co-60 4.08E+01 1.81E+03 2n-65 1.99E+01 6.43E+02 i
Sr-89 6.10E+01 1.92E+03 Sr-90 2.36E+03 7.44E+04 1
Zr-95/Nb-95 3.77E+00 1.24E+02 Ru-103 1.02E+01 3.22E+02 Ag-110m 3.63E+01 1.22E+03 Sb-224 6.95E+00 2.32E+02
{
I I-131 4.19E+02 -
1.32E+04 1-133 6.29E+00 1.98E+02 Cs-134 8.52E+01 2.83E+03 Cs-136 4.71E+00 1.52E+02 i
Cs-137 8.71E+01 2.97E.03 Ba-140/! a-140 1.44E+00 4.60E+01 i
Ce-141 9.75E-01 3.10E+01 Ce-144, 2.10E+01 6.65E+02
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3.0 OFF-SITE DOSE CALCULATION METHODS Chapter 3 provides the basis for plant procedures that the plant operator requires to meet the Dose Radiological Effluent Technical Specifications (hereafter called Dose RETS). A simple, conservative method (called Method I) is listed in Table 1.1-1 for each of th's nine doses required by the Dose RETS.
Each of the nine Method I ecuations 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, j
data, and reference material with which the operator can calculate the needed doses.
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3.1 Introductory Concepts I
The Radiological Effluent Technt:a1 Specifications (RETS) either limit dose or dose rate.
The term "Dose" for ingested or inhaled radioactivity means the dose commitment, measured in mrem, which results from the exposure i
to radioactive materials that, because of uptake and deposition in the body, will continue to expose the body to radiation for some period of time after the source of radioactivity is stopped.
The time frame over which the dose commitment is evaluated is 50' years.
The phrases "annual Dose" or "Dose in one year" then refers to the fifty-year dose commitment from one year's worth of releases.
"Dose in a quarter" similarly means a fifty-year dose commitment from one quarter's releases.
The term "Dose," with respect to external exposures, such as to noble gas clouds, refers only to the doses received during the actual time period of exposure to the radioactivity released from the plant. Once the source of the radioactivity is removed, there is no longer any additional accumulation to the dose commitment.
Gaseous effluents fro,m the plant are also controlled such that tha maximum "dose rates" at the site boundary at any time are limited ~to 500 mree'yr to the whole body or 3000 mrem /yr to the skin.
The annual dose Ilmits are the doses associated with the concentrations of Appendix B. Table II, Column 1 of 10CFR Part 20 (100FR20.106(a)).
The use of the annual dose 1
limits embodied in 10CFR Part 20 as plant "dose rate" values (to be applied at any time consistent with the capabilities of the monitoring instrumentation to determine) provides reasonable assurance that radioactive material discharged i
in gaseous ' effluents will not result in the exposure of member (s) of the pubile either within or outside the site boundary to annual average l
concentrations exceeding the federal regulations.
l
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It should also be noted that a dose rate due to noble gases that exceeds, for a short time period (less than one hour in duration), the i
equivalent 500 mrem / year dose rate limit stated in Technical Specification
(
3.11.2.1, does not necessarily by itself constitute a Licensee Event Report
]
(LER) under 10CFR Part 50.73 unless it is determined that the air l
concentration of radioactive effluents in unrestricted areas has also exceeded
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O e
k-two times MPC when averaged over one hour (four-hour notification per 10CFR50.72, and 30-day LER per 10CFR50.73).
l The quantitles D and D are introduced to provide calculable quantitles, related to cff-site dose or dose rate which demonstrates compliance with the RETS.
1 The dose D is the quantity calculated by the Chapter 3 dose equations.
The D calculated by "Method I" equations is not necessarily the actual dose received by a real individual but usually provides an upper bound for a given release because of the conservative margin butit into the dose factors and the j
selection and definition of critical receptors.
The radioisotope specific dose factors in each "Method I" dose equation represent the greatest dose to any organ of any age group accounting for existing or potential pathways of t
exposure.
The critical receptor assumed by "Method I" equations is typically a hypothetical individual whose behavior - in terms of location and intake -
{
results in a dose which is expected to be higher than any real individual, j
Hethod !! allows for a more exact dose calculation for real individuals, if necessary, by considering only existing pathways of exposure, or actual concurrent meteorology with the recorded release.
D is the quantity calculated in the Chapter 3 dose rate equations.
It is calculated using the plant's effluent monitoring system reading and an annual average or long-tert, atmospheric dispersion factor.
If plant release rates were such that 4 0 equal to the Technical Specification (3.11.2.1) value was continued for one year, the annual dose limits of 10CFR20 would be reached. However, since manimum allowed release rates and the resulting dose rates in the range of the Technical Specification limits are very infrequent, and are typically of short time duration, this approach of limiting dose rates t
equivalent to the annual dose limits then assures that 10CFR20.106 limits on an annual average air concentration in unrestricted areas will be met.
Each of the methods to calculate dose or dose rate are presented in separate sections of Chapter 3, and are summarized in Table 1.1-1.
Each Revisten 6 - 2/18/85 Approved By:4f
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5008R/12.332 3-3
i method has two levels of complexity and conservative margin and are called
~
Method I and Method II. Method I has the greatest margin and is the simplest; generally a linear equation. Method II is a more detailed analysis which allows for use of site-specific factors and variable parameters to be selected to best fit the actual release. Guidance is provided, but the appropriate margin and depth of analysis are determine ( in each instance at the time of analysts under Method !!.
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. _ _g 3.2 Method to Calculate the Total Body Dose from Liould Releases Technical Specification 3.11.1.2 limits the total body dose commitment to a Member of the Pubile from radioactive material in liquid effluents to 1.5 mrem per quarter and 3 mrem per year.
Technical Specification 3.11.1.3 requires liquid radwaste treatment when the total body dose estimate exceeds 0.06 mrem in any 31-day period.
Technical Specification 3.11.4 limits the total body dose comm1tment to any real member of the public from all station sources (including liquids) to 25 meem in a year. Dose evaluation is required at least once per 31 days.
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 11guld release from the plant.
Use Method II if a more accurate calculation of Total Body Dose is needed (i.e., Method I indicates the dose is greater than the limit), or if Method I cannot be applied.
To evaluate total body dose for Specification 3.11.1.3 add the Total Body Dose from today's expetted releases to the Total Body Dose accumulated l
for the time period of interest.
l 3.2.1 Method I The total body dose from a liquid release is:
Dtb
- K I 0 OI'itb (Eq. 3-1) 1 I
(mrem) where:
Site-specific tot'al body dose factor (mrem /C1) for liquid DFL tb release.
See Table 1.1-7.
Total activity (Curies) released to liquids of radionuclide Qg "1" during period of interest.
For 1 - Fe55, Sr89, Sr90, or H3, use the best estimates (such as the mo t racent measurements).
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5008R/12.332 3-5
366/Fo; where Fd is the average (typically monthly X
=
average) dilution flow of the Deerfield River below Sherman 3
Dam (in ft /sec).
If Fd etnnot be obtained or F 15 greater than 366 K can be assumed to equal 1.0.d The value 366 is the ten-year minimum monthly average 3
Deerfield River flow rate below Sherman Dam (in ft /sec).
Equation 3-1 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1.
Liquid releases to the circulating water pathway to Sherman l
Reservoir, or to the west storm drain pathway to the Deerfield River, and 2.
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 exceeds the liett or if a m re exact calculation is required, then Method II should be applied.
Method II consists of the models, input data and assumptions +1n Regulatory Guide 1.109, Rev. 1 (Reference A), except where site-specific models, data or assumptions are more appitcable.
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 6nd 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 Specifications which limit off-site total body dose from Itquids (3.11.1.2, 3.11.1.3, and 3.11.4) have been met for releases over the appropriate periods.
These Technical Specifications are based on design objectives and standards in 10CFR50 and Revision 6 - 2/18/88 Approved By: /
O 5008R/12.332 3-6
0 4
40CFR190.. Technical Specification 3.11.1.2 is based on the ALARA design objectives in 10CFR50, Appendix ! Subsection II A.
Technical Specification 3.11.1.3 is an "appropriate fraction", determined by the NRC, of that design objective (hereafter called the Objective).
Technical Specification 3.11.4 is based on Environmental Standards for Uranium Fuel Cycle in 40CFR190 (hereafter called the Standard) which applies to direct radiation as well as liquid and gaseous effluents. Method I applies only to the liquid contribution.
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 substantially underestimated", (10CFR50, Appendix I).
The definition, below, of a single "critical receptor" (a hypothetical individual whose behavior results in an unrealistically high dose) provides part of the conservative margin to tne calculation of total body dose in Method I.
Method II allows that actual individuals, with real i
behaviors, be taken into account for any given release.
In fact, Method I was based on a Method II analysis for the critical receptor and annual average 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 in mrem /C)) for a 1 curie release of each radionuclide in 11guld I
effluents was derived.
The base case analysis uses the methods, data and assumptions in Regulatory Guide 1.109 (Equations A-3, A-7, A-13 and A-16, l
Reference A).
Tables 3.2-1 rnd 3.2-2 outline human consumption and environmental parameters used in the analysis.
It is assumed that the j
critical receptor fisher celow Sherman Dam and eats the fish caught from this location and consumes leafy vegetables and produce from a farm which is irrigated with water from the Deerfield River below Sherman Dam.
Is is also h
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4 assumed that the critical receptor drinks milk and eats meat from cows who drink water from the Deerfield River below Sherman Dam and eat silage from the irrigated farm above.
The model is conservative because no real individual is likely to have that critical combination of exposures. A real individual would have cnly one or two pathways of exposure. A plant discharge flow rate-3 of 308 ft /sec was used with a mixing ratto of 0.84.
For any liquid release, during any period, the increment in annual average total body dose from radionuclide "1" is:
- 1tb
- 0 0FLitb 1
i i
where OFL is the total body dose factor for radionuclide "1" and Qg is itb the activity of radionuclide "i" released in curies.
t Method I is more conservative than Method II because it is based on the following reduction of the base case.
The dose factors. OFlitb, used in l
I He,thod I were chosen from the base case to be the highest of the four age groups for that radionuclide.
In effect, each radionuclide is conservatively l
represented by its own critical age group, i
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2 T'~s Revision 6 - 2/16/88 Approved By:
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Table 3.2-1 4
Environmental Parameters for liquid Effluents at Yankee Rowe (Derived fron Reference A) l POTABLE A00A11C SHORELINE F000 GROMN WITH CONTAMINATED WATER VARIABLE WATER F000 ACTIVITY VEGETABLES LEAFY VEG.
MEAT CON MILK GOAT MILK
- MP Mining Ratio 0.84 0.84 0.84 0.84 0.84 0.84 0.84 TP Transit Time (HRS) 12.0 24.0 0.0 0.0 0.0 480.0 48.0 48.0-j 2
YV Agricultural (KGfM )
2.0 2.0 2.0 2.0 2.0 Productivity 2
P Soll Surface (KG/M )
240.0 240.0 240.0 240.0 240.0 Density 2
IRR Irrigation Rate (L/M /HR) 0.15 0.15 0.15 0.15 0.15 i
TE Crop Exposure (MRS) 1440.0 1440.0 1440.0 1440.0 1440.0 Time TH Holdup Time (HRS) 1440.0 24.0 2160.0 2160.0 2160.0 QAW Water Uptake Rate (L/D) 50.0 60.0 8.0 for An1 mal Or reed Uptake Rate (KG/D) 50.0 50.0 6.0 for Animal 1
location of Mone Below Below Below Below Below 8elow None i
Critical Individual Sherman Sherman Sherman Sherman Sherman Sherman Dam Dam Dam Das Dam Dam
- Pethway is not included in Method I.
It is listed for information purposes and the possible use in a Method II.
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Table 3.2-2 Age Specific Usage Factors for Various Llouid Pathways at Rowe (From Reference A. Table ~E-5.
Zero where no pathway exists)
AGE VEG.
LEAFY MILK MEAT FISH INVERT.
POTABLE S h0:.i..
VEG.
WATER (KG/YR)
(KG/YR)
(LITER /YR)
(KG/YR)
(KG/YR)
(KG/YR)
(LITER /YR)
(hE',
Adult 520.00 64.00 310.00 110.00 21.00 0.00 0.00 10.C Teen 630.00 42.00 400.00 65.00 16.00 0.00 0.00 67.:
Child
$20.00 26.00 330.00 41.00 6.90 0.00 0.00 1 4. '. -
Infant 0.00 0.00 330.00 0.00 0.00 0.00 0.00 0.(:
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l 3.3 Method to Calculate Maximum Organ Dose from Liould Releases Technical Specification 3.11.1.2 limits the maximum organ dose j
e commitment to a Member of the Public from radioactive material in liquid j
I effluents to 5 mrem per quarter and 10 mrem per year.
Technical Specification 3.11.1.3 requires 11guld radwaste treatment when the maximum organ dose estimate exceeds 0.2 mrem in any 31-day period.
Technical Specification l
3.11.4 limits the maximum organ dose commitment to any real mcmber of the
(
public from all station sources (including liquids) to 25 mrem in a year except for the thyrold, which is limited to 75 mrem in a year.
Dose evaluation is required at least once per 31 days.
If the liquid radwaste i
treatment system is not being used, dose evaluation is required beforo each release.
i Use Method I first to calculate the maximum organ dose from a liquid release from the plant.
1
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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 1 l
- i cannot be applied.
7 f
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I To evaluate the maximum organ dose for Specification 3.11.1.3, add the
- [
organ dose from the expected releases to the organ dose accumulated for the
(
time period of interest.
l 3.3.1 Methed I l
The maximum organ dose from a 11guld release is:
f (Eq. 3-2)
Kf0DFL D
1 gg organ (arem) where:
Site-specific maximum organ dose factor (mrem /C1) for a DFiimo Itquid release. See Table 1.1-7.
Approved Byy4,i/-.
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500SR/12.332 3-11
Total activity (Curies) released to liquids of radionuclide I
.Qg "1" during period of interest.
For i. Fe55, Sr89, Sr90, I
or H3, use the best estimates (such as the most recent measurements).
366/F ; where Fd is the average (typically monthly K
d average) dilution flow of the Deerfield River below Sherman 3
Dam (in ft /sec).
If Fd cannot be obtained or Fd Il greater than 366, K can be assumed to equal 1.0.
The value 366 is the ten-year minimum monthly average 3
Deerfield River flow rate below Sherman Dam (in ft /sec).
Equation 3-2 can be applied under the following conditions (otherwise, justify Method I or consider Method !!):
I 1.
Liquid releases to the circulating water pathway to Sherman Reservoir, or to the west storm drain pathway to the Deerfield River, and 2.
Any continuous or batch release over any time period.
3.3.2 Method 11 If Methc: Icann$tbeapplied,oriftheMethodIdoseexceedsthe limit or if a more enact calculation is required, then Method !! should be applied. Method !! 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 appitcable.
The case case analysts.
documented telow, is a good example of the use of Method II.
It is an l
l acceptable starting point for a Method !! analysis.
1 1
3.3.3 Basis for Method I 1
l 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 maximum organ dese parallel the total body dose methods (see Section 3.2.3).
Only the differences are presented bere.
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o For any liquid release, during any period, the increment in annual average dose from radionuclide "1" to the maximum organ is:
, 40%. Qg DFL,
g where DFL is the maximum organ dose factor for radionuclide "1" and Qg_
is the activity of radionuclide "1" released in curies.
The dose factors DFL used in Method I were chosen from the base gg case to be the highest of the set of seven organs and four age groups for each radionuclide.
This means that the maximum effect of each radionuclide is conservatively represented by its own critical age group and critical organ.
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3.4 Method to Calculate the_ Total Body Dose Rate From Noble Gases i
Technical Specification 3.11.2.1 limits the dose rate at any time to the total body from noble gases at any location at or beyond the site boundary l
equal to or less than 500 mrem / year. By limiting the maximum Otb to a rate equivalent to no more than 500 mrem / year, assurance is provided that the total body dose accrued in any one year by any member of the general public will be
[
Iess than 500 mrem in accordance with the annual dose limits of 10CFR Part 20 to unrestricted areas.
t r
Use Method I first to calculate the total body dose rate from the peak release rate via the plant vent stack. Method I applies at all release rates, i
Use Method !! if Method I predicts a dose rate greater than the j
Technical Specification limit (i.e., use of actual meteorology over the period 3
of interest) to determine if, in fact, Technical Specification 3.11.2.1 had i
actually been exceeded during a short time interval.
I Compliance with the dose rate limits for noble gases is continuously j
?
demonstrated when effluent release rates are 6elow the plant vent stack noble l
gas activity monitor alarm setpoint by virtue of the fact that the aiarm
[
I setpoint is based on a value which corresponds to the off-site dose rate limit i
of Technical Specification 3.11.2.1. or a value below 1t.
j f
l Determinations of dose rates for compilar.ce with Technical
[
Specification (3.11.2.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Specification l
3.11.2.1 is unsuccessful, or as required by the action to Technical Speelfication Table 3.3.9 when the stack noble gas monitor is inoperable.
t i
3.4.1 Method I l
6 The Total Body Dose Rate due to noble gases can be determined as fo11ons:
Revision 6 - 2/18/88 Approved By:Af d
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c b ("
) = 7.83 h 0FB (Eq. 3.3) tb g
g where:
t In the case of noble gases, the release rate from the plant Qg vent stack (pC1/sec) of radidnuclide "1".
The release rate at the plant vent stack is based on the measured radionuclide distributton in the off-gas during plant operation, and the recorded total gas effluent count rate from the St.ck noble gas activity monitor.
The release rate at the stack can also be stated in the following equation:
hd M F
g (Eq. 3-10)
I
$.( ) (cpm) (pCl/cc) gic_)
l sec epm se; j
where:
Plant vent stack monitor count rate (cre).
M
=
l Appropriate plant vent stack monitor detector counting Sg efficiency (com/(pct /cc)).
l Plant vent stack flow rate (cc/sec).
F
=
i Fraction of the release which is nuclide "1".
This fraction i
di can be based on the last measured value o' nuclide "1" witt i
respect to the total noble gas activity released at the PV5.
l It can also be based on the fraction of nucitde "i" in the i
primary coolant with respect to the total noble gas primary coolant activity, j
total body dose f actor (see 'iable 1.1-2) i DFBt Durinl ceriods (beyond the first two days) when the plant is shutdown and no radioattivity release rates can be measured at the PV5 Xe-133 may be used as the referenced radionuclide to determine off-site dose rate and monitor setpoints. Alternately, a relative radionuclide "1" mix fraction (f ) may be taken from Table 5.2-2 as a function of time af ter shutdown, anc g
substituted in place of d in Equation (3-10) above to determine the f
g relative fraction of each noble gas potentially available for release to the
/ s Revision 6 - 2/18/88 Approved By: A b / g 5008R/12.332 3-15
6 t M.. Ju.t prior to plant startup, the.nonttor alarm setpoints, should be based on Xe-138 as representing the most prevalent high dose factor noble gas expected to be present shortly after the plant returns to power. Monitor alarm setpoirits which have been determined to be conservative under any plant conditions may be utili::ed at any time in lieu of the above assumptions.
Equation 3-3 can be applied under the following conditions (otherwise, justify Method I or con.ider Method II):
1.
Normal. operations (not emergency event),
2.
Noble gas releases via the plant vent stack to the atmosphere.
l 3.4.2 Method II If Method ! cannot be f.pplied, or if the Method I dose exceeds the limit or if a more exact calculation is required, then Method II may be applied.
Method 11 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 r'or Method _I This section serves four purposes:
(1) to docum ht that Method I complies with appropriate NRC regulations, (2) to define the word "rate" as used in the Technical Specificatica, (3) to provide background and trainir,g information to Method I users, and (4) to provide an introductory user's guide to tiethed II.
Method I may be used to show that the Technical Specification which limit total body dose rate from noble gases released to the atmosphere (Technical Specification 3.11.2.1) has been met for the peak noble gas release rate.
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o Technical Specification 3.11.2.1 ensures "that the doses... at and beyond the SITE BOUNDARY from gaseous effluents... will be within the annual dose limits... associated with the concentrations of 10CFR Part 20, Appendix B, Table II, Column 1" (Technical Specification 3/4.ll.2.1-Basis).
Toose "Maximum Permissible Concentrations" for air in unrestricted areas, called ura MPC
, cannot be exceeded if this Technical Specification is met.
Hence the requirements of 10CFR20.106(d) are met.
Because the plant has never ura approached even a small fraction of HPC limits Technical Specification 3.11.2.1 was given greater conservative margin by the NRC.
It additionally restricts release rate monitor readings to the level at which the plant could operate continuously and not exceed the annual dose limit.
The annual total body dose limit is 500 mrem (from NBS Handbook 69, Reference G Page 6), which isthebasisfortheMPCfA limits.
Exceeding the annual average total body dose rate could result in plant shutdown, especially if the operators cannot take action to reduce the peak release rate.
Method I was derived from Regulatory Guide 1.109 as follows:
T 4
D - 3.17 x 10 X/Q S IQ CFB p
g I
was derived by combining Equations B-4 and B-5 from Regul6 tory Guide 1.109, D
assuming X/Q - )/Q for noble gases, and some simplification in the notation.
Assumir.g that D inite D gyjg)y/(X/Q),andthatb l
T tb " finite
.Q(pCi/sec).31.54/Q(C1/yr) we get:
b (mrem /yr) - 1 x 10 Sp (X/Q)Y hg 6
DFB tb g
substituting:
5
= 1.0 (shielding factor) i 7
(X/Q)Y = Long-Term Average Gamma Dilution Factor 3
7.83 x 10-6 (sec/m )
Revision 6 - 2/18/88 Approved By:fti, As 500BR/12.332 3-17
i h
- Release rate of noble. gas "1" (pC1/sec) 3 gives:
btb (mrem /yr) = 7.83 I h (pC1/sec) DFB, (Eq. 3-3) j i
MethodIIcannotprovidemuchextrarealismbecauseb is already tb based on several factors which make use of current plant parameters.
- However, should it be needed, the dose rate analysis for critical receptor can be performed making use of current meteorology during the time interval of recorded peak release rate in place of the. default atmospheric dispersion factor used in Method I.
I l
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3.5 Method to Calculate the Skin Dose Rate from Noble Gases Technical Specification 3.11.2.1 limits the dose rate to the skin from l
noble gases at any locat'on at or beyond the site boundary to 3,000 mrem / year.
By limiting the maximum D to a rate equivalent to no more than sk 3,000 mrem / year, assurance is provided that the skin dose accrued in any one year by any member of the general public is much less than 3,000 mrem.
I Use Method I first to calculate the skin dose rate from the peak relea3e rate via the plant vent stack.
Method I applies at all release rates.
Use Method I! if Method I predicts a dose rate greater than the Technical Specification limits.(i.e., use of actual meteorology over the period of interest) to determine if, in fact, Technical Specificat'Sn 3.8.E.1 had actually beEn exceeded during a short time interval.
i Compliance with the dose rate limits for noble gases is continuously j
demonstrated when effluent release rates are below the plant vent stack noble gas activi';; monitor alarm setpoint by virtue of the fact that the alarm setpoint is based on a value which corresponds to the off-site Technical l
Specification dose rate limit, or a value below it.
Determinations of dose rate for compliance with Technical Specification (3.11.2.1) are performed when the effluent monitor alarm setpoint is exceeded and the corrective action required by Specification 3.11.2.1 is unsuccessful, or as required by the notations to Technical l
Specification Table 3.3.9 when the stack noble gas monitor is inope.able.
l i
3.5.1 Method I The Skin Dose Rate due to noble gases is:
bskin (mrem /yr) =
I h DF (Eq. 3-4) g i
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M A
5008R/12.332 3-19
'(
l
6.
where:
ht In the case of noble gases, the release rate from the plant vent stack (pC1/sec) of each radionuclide "1".
The release rate at the plant vent stack is based on the measured vadionuclide distribution in the off-gas during plant operation, and the recorded total gas effluent count rate from the stack noble gas activity monitor.
The release rate at the i
stack can also be stated in the following equation:
6=d M F
g (Eq. 3-10) uC1, ( ) ggp,) guC1/cc) (1) sec cpm see where:
Plant vent stack monitor count rate (cpm).
M
=
g Appropriate plant vent stack monitor detector. counting S
efficiency (cpm /(pC1/cc)).
lant vent stack flow rate (cc/sec).
F
=
Fraction of the release which is nuclide "i".
This fraction dj
=
can be based on th( last measured value of nuclide "1" with i
respect to the total 4oble gas activity released at the PVS.
It can also be based on '. fraction of nuclide "1" with l
respect to the total noble gas primary coolant activity.
skin dose factor (see Table 1.1-2)
DFj
=
Duriig periods (btyond the first two days) when the plant is shutdown and no radioactivity release rates can be measured at the PVS. Xe-133 may be used as the referenced radionuclide to determine off-site dose rate and monitor setpoints. Alternately, a relative radionuclide "1" mix fraction (f ) may be taken from Table 5.2-2 as a function of time after shutdown, and g
substituted in place of d in Equation (3-10) above to determine the g
relative fraction of each noble gas potentially available for release to the total. Just prior to plant startup, the monitor alarm setpoints should be based on Xe-138 as representing the most prevalent high Jose factor noble gas expected to be present shortly after the plant returns to power. Monitor Reelston 6 - 2/18/88 Approved By:,4f
/
(
5008R/12.332 3-20
]
alarm setpoints which have been determined to be conservative under any plant conditions may be utilized at any time in lieu of the above assumptions.
Equation 3-4 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1.
Normal operations (not emergency event),
2.
Noble gas releases via the plant vent stack to the atmosphere.
l 3.5.2 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 11 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 appilcable.
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 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.
The methods to calculate skin dose rate parallel the total body dose rate methods in Section 3.4.3.
Only the differences are presented 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 3.11.2.1) has been met for the peak noble gas release rate.
The t.nnual skin dose limit is 3,000 mrem (from NBS Handbook 69 Reference G, Pages 5 and 6, is 30 rem /10), wh'ch is the basis for the MPC["limits.
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5008R/12.332 3-21
Hethod.I was derived from Regulatory Guide 1.109 as follows:
0 - 3.17 x 104 [X/Q 't.11 S
I Q DF{ + X/Q I Q OFS )
3 p
g g
i 1
was de.1vsd by combining Equations B-4, B-5, and B-7 from Regulatory Guide D
1.109, assuming that X/Q - X/Q for noble gases, and making some simplifttations in notation. Assuming that Dfinite " 0 03
' 03 and that Osk - 05
- Q(pC1/sec)
- 31.54/Q(C1/yr) yleids x (1 x 10b (X/Q)Y
{ h gp bskin (mrem /yr) - 1.11 x Sp i
1
+ 1 x 106 (X/Q)
I h DFS 3
g 1
where:
3 (X/Q)Y 7.83 x 10-6 sec/m X/Q
- 2.39 x 10-5 3,gj,3 S
= 1.0 (shielding factor) l F
Substituting gives l
1 6,ggn (mrem /yr>
s.69 I h DF{ + 23.9 x 10 I h, oFS.
3 g
1 1
I hg (8.69 0FJ + 23.9 0FS ]
g 1
Define:
DF -8.690F{+23.90FS g Then:
bskin (mrem /yr) -
I hg DF (Eq. 3-4) 1 l
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_. - =
~
3.6 Method to Calculate the Critical Organ Dose Rate from 131-I, 3-H and Particulates with T Greater Than 8 Days jfp Technical Specification 3.11.2.1 limits the dose rate to any organ from 131-1, 3-H and radionuclides in particulate form with half lives greater than 8 days to 1500 mrem / year to any organ.
The peak release rate averaging time in the case of lodines and particulates is commensurate with th6 time the iodine and particulate samplers are in service between changeouts (typically a week).
By limiting the maximum 1
b to a rate equivalent to no more than 1500 mrem / year, assurance is en provided that the critical organ dose accrued in any one year by any member of the general public will be less than 1500 mrem.
l Use Method I first to calculate the 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 Method I predicts a dose rate greater than the l
Technical Specification limits (i.e., use of actual meteorology over the period of interest) to determine if, in fact, Technical Specification 3.11.2.1 had actually been exceeded during the sampling period.
I 3.6.1 Method I The critical organ dose rate can be determined as follows:
O,-
IQg DFG Uq. 3-5) g jgo i
mrem
}g mrem-see yr sec pCi-yr Revision 6 - 2 18/88 Approved By: # [
M
_s 5008R/12.332 3-23
c where:
Stack activity release rate determination of radionucliue Og "1"
(iodine, tritium, tid part'culates with half-lives
'sreat.'r than 8 days) in pCi/sec.
For i = Sr89, Sr90, or H3, usa the best estimates (stah as most recent measurements).
Site-specific critical organ dose rate factor ("I'**S'C) for DFG'CU
=
I pCi-yr a oiseous release.
(See Table 1.1-8.)
Equation 3-5 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1.
Normal operations (not emergency event),
2.
Tritium, iodine, and particulate releases via the plant vent.
I stack to the atmosphere As an alternative to performing Method I calculations, compliance with the critical organ dose rate limit of 1,500 mrerr/ year in TecMical Specification 3.11.2.1.b can be shown by two methods.
They are:
a comparison of the measured 1-131 release rate to determine if it is at or below an inspection limit of 0.0125 pC1/sec, or a concentration limit in the PVS equivalent to:
C I-131 ( C1/cc) = (2.65 x 10-5)/F (cfm)
PVS where:
F = average PVS flow rate measured during the sampling interval.
This results from the fact that I-131 is the controlling nuclide with respect to any critical organ dose, and the select 2d inspection limits represent approximately ten percent (10 percent) o? the 1,~00 mrem / year dose rate limit.
Measured values greater than the inspection limits should be evaluated by equation 3 5, or a Method II assessment.
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l[:
a 3.6.2 Method II
-If Method I cannot be applied, or if the Method I dose exceeds the Technical Specification limit or if a more exact calculation is required, then Method 11 may be applied.
Method II consists of the models, input data and assumptions in Regulatory Guide 1.109, Revision 1 (Reference A), except where site-specific models, data or assumptions are more applicable.
The base case analysis, documented below, is a prod example of the use of Method II.
It is an. acceptable starting point for a Method II analysis.
3.6.3 Basis for Method I This section serves four purposes:
(1) to document that Method '
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 dose rate methods in Section 3.4.3.
Only the differences are presented here.
Method I may be used to show that the Technical Specification which limits organ dose rate from Iodine-131, Tritium and radionuclides in particulate form wi'.h half lives greater than 8 days released to the j
atmosphere (Techrscal Specification 3.11.2.1) has been met for the peak Iodine-131, tr' tium, and particulate release rates.
Theequationforh is derived by modifying Equation 3-8 from go Section 3.9 as follows:
IQDM (Eq. 3-8)
D 3
ico eo i
N
- I'*
mrem yr Ci applying the conversion factor, 31.54 (Ci-sec/pci-yr) and converting Q to Q in pC1/sec yields Revision 6 - 2/18/88 Approved By M,d,
~
4 w
/
5008R/12.332 3-25 L
2 o
I 31.54 h DFG O
go g
gg, pCi mrem C
i mrem Ci-sec yr pC1-yr sec C1 Eq. 3-5 is rewritten in the form:
b I b DFGjg,
=
co g
i where:
DFG 31.54 DFGjen ggo mrem-sec mrem Ci-sec pC1-yr Ci pCi-yr Should Method II be needed, the analysis for critical receptor critical pathway (s) and the annual average dispersion coefficients may be performed with actual meteorologic and latest la'nd use census data to identify the location of those pathways which are most impacted by these types of releases.
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5008R/12.332 3-26
3.7 Method to Calculate the Gamma Air Dose from Noble Gases Technical Wecification 3.11.2.2 limits the gamma dose to air from noble gases at any lor;ntion at or beyond the site boundary to 5 mrad in any quarter and 10 mraJ in any year. Dose evaluation is required at least once per 31 days.
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 ir(mrad) = 0.25 IQ DF}
(Eq. 3-6) j I
where:
Qg
- total activity (Curles) released to the atmosphere via the plant vent stack of.each radionuclide "l" during the period of interest.
OF{=gammadosefactortoairforradionuclide"1".
See Table 1.1-2 Equa+1on 3-6 can be applied under the following conditions (otherwise justify Metnod I or consider Method II):
1.
Normal operations (not emergency event),
2.
Noble gas releases via the plant vent stack to the atmosphere.
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5008R/12.332 3-27 T-
3.7.1.1 Ground Level Releases For ground level releases, the gamma air dose is:
grd (mradi. (1.23 x 10~4) (QXe-133 equivalent)
(Eq. 3-6.1)
O where:
the Xe-133 equivalent of all the noble gases in QXe-133 equivalent the release (curies), and is based on the dose conversionfactors(DF{}aslistedin Table 1.1-2 of the 00CM.
3.7.2 Method II If M.ethod 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 Reguistory 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.
3.7.3 Basis for Method I This section serves three purposts:
(1) to document that Method I cornplies 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 effluents (3.11.2.2) has been met 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 gamma air dose at unrestricted area locations.
Revision 6 - 2/18/88 Approved By: 4G; 5008R/12.332 3-28
)
Exceeding the Objective does not immediately limit plant operation but requires a report to the NRC within 30 days.
For any noble gas release, in any period, the dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109 with the added assumption that Dfinite - 0 U/Q)Y [X/Q):
/
3 D tr(mrad) - 3.17 x 104 ( h h ) (X/Q)Y (sec/m )
Qg(C1)DF{(y
)
where:
(X/Q)Y - long-term average gamma dilution factor 3
- 7.83 x 10-6 (sec/m )
Qg
- number of curies of noble gas "1" released which leads to:
D ir(mrad) = 0.25 IQg(C1)DF{
(Eq. 3-6) i The main difference between Method I and Method II is that Method II l
would allow the use of actual meteorology to determine (X/Q)Y rather the-use the maximum long-term average value obtained.for the time period frorn 1/81 through 12/85.
The gamma-air dose from a ground level release is determined by using tht same Regulatory Guide 1.109 equation to derive Equation 3-6 above.
The only differences are:
(X/Q)Y,
1,1 x 10-5 3,ej,3, which is tne long-term average ground level (X/Q)Y based on the time period from May 1977 through April 1982.
OXe-133 equivalent - the Xe-133 equivalent of all the noble gases in the release (curies).
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4 DF{=DFXe-133 - 3.53 x 10 (hc-y
)
btained from Table 1.1.2, to to account for the release being expressed in terms of Xe-133 equivalent.
Substituting the above into the Regulatory Guide 1.109 general equation gives:
grd (mrad) = 1.23 x 10-4Q (Eq. 3-6.1)
O Xe-133 equivalent.
j I
e 4
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w r
3.8 Method to Calculate the Beta Air Dose from Noble Gases Technical Specification 3.11.2.2 limits the beta dose to air from 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 during the period. Method I applies at all dose levels.
Use Method II If a move accurate calculation is needed of if Method I cannot be applied.
3.8.1 Method I The beta air dose from plant vent stack releases is:
= 0.76 { ODFf (Eq. 3-7)
Dfir(mrad) g I
where:
CFf=betadosefactortoairforradionuclide"i".
See Table.l.1-2 O
= total activity (Curies) released to the atmosphere via the plant g
vent stack of each radionuclide "1" during the period of interest.
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 vent stack to the atmosphere.
L O
l 5008R/12.332
- 3-31 1
a 3.8.1.1 Ground Level Releases For ground level releases, the beta air dose can be determined by using Equation 3-7.
Equation 3-7 results in doses that are approximately 10 percent more conservative than calculating releases using ground level methodology.
3.8.2 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 Regulato'y 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.
3.8.3 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 calculate beta air dose parallel.
the gamma air dose methods in Section 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 effluents (3.11.2.2) has been met 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.
ExceLding the Objective does not immediately limit plant operation but requires a report to the NRC within 30 days.
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M 5008R/12.332 3-32 A
for any noble gas release,'in any period, the dose is taken from Equations B-4 and B-5 of Regulatory Guide 1.109:
B 0
D g7(mrad) - 3.17 x 104 (X/0) I Qg DF i
substituting X/Q = 2.39 x 10 sec/m3 He have 0
0 D 1r(mrad) - 0.76 IQg(C1)DF (Eq. 3-7) 1
\\
t
(
t Revision 6 - 2/18/88 Approved By:42 5008R/12.332 3-33
/
3.9 Method to Calculate the Critical Organ Dose from Tritium. Iodines and Particulates Technical Specification 3.11.2.3 limits the critical organ dose to a Member of the Public from radioactive Tritium, Iodine-131, and particulates with half-lives greater than 8 days in gaseous effluents to 7.5 mrem per quarter and 15 mrem per year.
Technical Specification 3.11.4 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.
l 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.
l 3.9.1 Method I The critical organ dose from a gaseous release is:
Qg 0FG (Eq. 3-8)
D
=
jgo en (mren) where:
Total activity (Curies) released to the atmosphere of Q) radionuclide i during the period of interest.
For 1 -
Sr-89, Sr-90, or H-3, use the best estimates (such as the most recent measurements).
Site-specific critical organ dose factor (mrem /C1) for a OFG co i
gaseous release.
See Table 1.1-8.
Equation 3-8 can be applied under the following conditions (otherwise, justify Method I or consider Method II):
1.
Normal operations (not emergency event),
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5008R/12.332 3-34
2.
Iodine, tritium, and particulate releases via the plant v~nt stack l
e to the atmosphere, and 3.
Any continuous or batch release over any time period.
3.9.2 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.
3.9.3 Basis for Method I This section serves three purposes:
(1) to document that Method I complies with appropriate NRC re'gulations, (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 Specifications which limit off-site organ dose from gases (3.11.2.3 and 3.11.4) have been met for releases over the appropriate periods.
These Technical Specifications are based on Objectives and Standards in 10CFR and 40CFR.
Technical Specification 3.11.2.3 is based on the A1. ARA Objectives in 10CFR50, Appendix I, Subsection II C.
Tec'hnical Specification 3.11.4 is based on Environmental Standards for Uranium fuel Cycle in 40CFR190 (hereafter called the Standard) which applies to direct radiation as well as 11guld and gaseous effluents. These methods apply only to iodine, tritium, and particulates in gaseous effluents contribution.
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5008R/12.332 3-35
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 such 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 3.11.2.3 and 3.11.4 annual average dilution 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 base case, the dose impact to the critical receptor in the form of dose factors, OFGggg (mrem /C1) of I curie release of each lodine, tritium, and l
I
, particulate radionuclide to gaseous effluents was derived.
Then Method I was determined using simplifying 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 "maximum site boundary dilution factor location" as defined in Section 3.10.
For any gas release, during any period, the dose from radionuclide "1" is:
Ogg,= OFGgen g Q
where DFG is the critical dose factor for radionuclide "1" and Qg is gen the activity of radionuclide "i" released in curies.
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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 DFG used in Method I were chosen from ggo 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 and critical organ.
T l
i t
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_ - =
Table 3.9-2
- ~
Environmental Parameters for Gaseous Effluents at Yankee Rowe (Derived from Reference A)
Vegetables Cow Milk Goat Milk
- Meat l
Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored YV Agricultural (Kg/M )
2.
2.
0.7 2.
0.7 2.
0.7 2.
]
2 Productivity 2
P Soll Surface (KG/M )
240.
240.
'240.
240.
240.
240.
240.
240.
Density T
Transport Time (HRS) 48.
48.
48.
48.
480.
480.
to User TB Soll Exposure (HRS) 131400.
131400.
131400.
131400.
131400.
131400.
131400.
131400.
Time TF Crop Exposure (HRS) 1440.
1440.
720.
1440.
720.
1440.
720.
1440.
l Time to Plume TH Holdup After (HRS) 1440.
24.
O.
2160.
O.
2160.
O.
2160.
Harvest OF Animals Daily (KG/ DAY) 50.
50.
6.
6.
50.
50.
Ferd i
FP Frcction of Year 0.50 0.50 0.50 on Pasture
' Pathway is not included in Method I.
It ts listed for information purposes, and the possible use in a Method II analysis.
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w l
5008R/12.332 3-39 e
A
_m
=-_.sL._
4 ur,._
L_
~'
Table 3.9-2 (continued)
Environmental Parameters for Gaseous Effluents at Yankee Rowe (Derived from Reference A) l Vegetables Cow Milk Goat Milk
- Meat Variable Stored Leafy Pasture Stored Pasture Stored Pasture Stored FS Frtction Pasture 1.
1.
1.
j when on Pasture i FG Fraction of Stored 0.76 V:g. Grown in Garden FL Fraction of Leafy 1.0 V29. Grown in Garden
' FI Fraction Elemental i
Iodine - 0.50 3
H Absolute (ge/M 3 Humidity - 5.6 i
~
1 i
4
{*PathwayisnotincludedinMethodI. It is listed for information purposes. and the possible use in a Method II analysts.
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b __
I 500RR/12.332 3-40 I
h.
h 3.10 Critical Receptors and Long-Term Average Atmospheric Dispersion Factors for Important Exposure Pathways The gaseous effluent dose equations (Method I) have been simplified by assuming an individual whose behavior and living habits inevitably lead to a higher dose than anyone else.
The following pathways of exposure to gaseous effluents as listed in Regulatory Guide 1.109 (Reference A) have been considered.
They are:
Direct exposure to contaminated air, Direct exposure to contaminated ground, Inhalation of air, Ingestion of vegetables, Ingestion of cow milk, and 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 dispersion factors; and Section 3.10.3 contains the resulting descriptions of the critical receptors and their dispersion factors as a function of exposure pathway.
3.10.1 Critical Receptors t
The most limiting site boundary location in which individuals are, or likely to be, located was assumed to be the receptor for all the gaseous pathways considered.
This provides a conservative estimate of the dose to an individual from existing and potential gaseous pathways for the Method I analysis.
l This point is the SSE sector, 800 meters.
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e 3.10.2 Yankee Atmospheric Dispersion Model The annual average dispersion factors are computed for routine (long-term) releases using the Yankee Atomic Electric Company's (YAEC) AEOLUS (Reference B) Computer Code.
AEOLUS produces the following annual average dispersion factors for each location:
g, nondepleted dispersion factors for evaluating ground level concentrations; (X/Q)D, depleted dispersion factors for evaluating ground level concentrations of iodines and particulates; T
B, effective gamma dispersion factors for evaluating gamma dose rates fro.n a sector-averaged finite cloud (multiple-energy undepleted, source); and O/Q, deposition factors for dry deposition of elemental radiolodines and other particulates.
The AEOLUS diffusion model is described in the AEOLUS Computer Code Manual (Reference B).
AEOLUS is based, in part, on the straight-line airflow model as discussed in Regulatory Guide 1.111 (Reference C).
One difference is that gamma dose rate is calculated throughout this OOCH using the finite cloud model presented in Meteorology and Atomic Energy 1968 (Reference H, Section 7-5.2.5).
That model is implemented through the definition (Reference B, Section 6) of an effective gamma dispersion factor, Y
X/0, and the replacement of X/Q in infinite cloud dose 1guations by the Y
X/Q.
1 i
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The other difference.is that the relatively narrow valley in which the plant sits is considered by the model. Wind channelling is assumed to occur in the sever, sectors which make up the valley.
The seven sectors are SSE, S.
SSH, SH, HSH, H, and HNH.
If a receptor locatiott is in one of the valley sectors, the contribution from the other six valley sectors are tveraged into the particular valley receptor.
This is done for distances greater than 500 meters from the primary vent stack where the valley effects are assumed to cause channelling.
3.10.3 Long-Term Average Dispersion Factors for Critical Receptors Actual measured meteorological data for the five-year period, 1/81 through 12/85, were analyzed to determine the locations of the maximum off-site atmospheric dispersion factors.
Each dose and dose rate calculation incorporates the maximum applicable off-site, long-term average atmospheric dispersion factor.
The values used and their locations are summarized in Table 3.10-1.
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TABLE 3.10-1
^
Yankee Nuclear Power Station Five-Year Average Atmospheric Dispersion Factors
- Dose to Critical Dose Rate to Individual Dose to Air Dan Total Body _
Skin Critical 0rgan__
_ Gamma _
Beta Thyroid X/0 depleted (y) 2.19E-06 2.19E-06 m
X/0undepleted(y) 2.39E-06 2.39E-06 6.
D/0 (
)
5.02E-08 5.02E-08 X/0Y(E) 7.83E-06 7.83E-06 7.83E-06 O
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3.lf Method _*o Calculate Direct Dose from Plant Operattoj Technical Specification 3.11.4 restricts the Dose to the whole body and any organ of any real member of the public from al' station sources (includinf direct radiation from the reactor and outside storage tanks which is called the Direct Dose) to the limit of 25 mrem in a year, except for the thyr, d which is limited to 75 mrem in a year. A determination of the need to conduct l,
a total dose evaluation is required at least every 31 days.
Use Method I first to calculate the Direct Oose contributton to the whole body and any organ as it is simpler to execute and more conservative than Method II.
Use Method II if a rnore accurate calculation of. rect Dose is needed or if Method I cannot be applied.
3.11.1 Method I The maximum contribution of Direct Dose to the whole body or to any organ is:
d - (0.057 + k ) T, 0.00087 (Eq. 3-9)
D r
where:
0.00087 = Conversioa factor (mrem!pR).
T,
. Length of exposure period in hours.
E Exposure rate at critical receptor from non-V.C. sources as r
measured or estimated for the perloo.
Equation 3-9 can be applied under the following conditions, otherwise justify Method I or consider Method II:
1.
Normal operations (not errergency event).
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k
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2.
All significant remaining sources are considered in E ' ""d r
3.
Any exposure period.
3.3.2 Mtthod y If Method I cannot be applied, or if the Method I dose exct.eds the limit or if a more exact calculation is required, then Method II should be applied. Method II consists of measurement and site-specific models, data and assumpticas.
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.
3.3.4. Basis for Method I ibis 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 t.i show that the Technical Specification that limits Direct Dose off-site (3.11.4) has been met for any exposure period.
Technical Specification 3.11.4 is based on the Standard (40CFR190) which applies to direct
'rces of radiation as well as liquid and gaseous effluents. Method I applies to the direct sources only.
Exceeding the Standard ooes 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 is developed below by reducing the "base case" (a Method II analysis) using conservative assumptions. The base case involves the choice of a critical receptor and the development of an exposure factor for the vapor container source, E
, (pR/ hour operation),
The critical receptor is the l
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nearest resident who lives 450m from the Vapor Container centerline in the NNW direction.
An occupency of 1.0 is assumed.
Theexposurefactork is developed below by extrapolating VC measurements made close to the plant, out to the critical receptor. All significant sources of direct radiation on-site are chtelded by buildings and tanks from the critical receptor with the exception of the Vapor Container and one of the 11guld waste storage tanks, that is Number TK-31.
The dose (mrem) to the critical receptor, D over the exposure period d
(in hours), T,, is related simplistically to the exposure rate from the Vapor Container in pR/ hour EVC, and the exposure rate from remaining sources, E, by the f ' towing equation:
r Dd - (EVC
- r) T, 0.00087 (Eq. 3.1-1)
What remains is to conservatively derive:
E and E VC r
The dose from the Vapor Container is due to fission and activatirn gases which build up in plant systems and the Vapor Container during operation.
Those sources decay or are ventilated through the Ventilation Exhaust Filtration System at the beginning of refueling outages.
This has to be done to allow worker access to the Vapor Container to reload the core.
The estimate of E is based on the extrapolation of VC measurements rnade during plant operation at the restricted area fence, E),
compared to background measurements made during a refuelfng outage after containment purge at the same locations E.
E is expected to remain b
ye constant over the years and so it can be estimated here as a function of exposure period.
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Although regular measurements of direct radiation are made near the critical receptor as part of the environmental surveillance program, the majority of the measured doses are due to natural background and fallout, variations in which entirely obscure plant contributions to dose.
Because they are closer to the sources, the measurement of direct radiation using TLDs at the restricted area fence can be extrapolated with greater net sensitivity (about 10 mR/ year).
However, the most sensitive method proved to be exposure rate measurements mr.de with a High Pressure Ionization Chamber, which had a htstory of ipr / hour plus or minus lpR/ hour 95 percent confidcnce interval for
.l exposure rates near the background rate for the procedure used (20 replicate measurements and periodic instrument checks).
This extrapolates to approximately a plus or minus Imrem/ year 95 percent confidence interval at the critical receptor, k
is estimated using the.following equation:
VC I
d kVC
- b)
(Eq. 3.11-2) where:
d. Distance to critical receptor from Vapor Container centerline.
1 d. Distance to exposure measurement from Vapor Container centerline.
o S
Exposure rate (pR/hr) measurement during plant operation.
k = Exposure rate (pR/hr) measurement during plant outage.
b k
is der ved from data collected in 1981 and presented in Table VC 3.11-1.
The mean value of EVC, for measurements at each of the nine TLD locations at the restricted area fence is 0.057 pR/ hour.
The mean value is used because it is insensitive to miscellaneous on-site sources which contribute to the measurements but not to th' dose at the critical receptor, k will have to be made from measurements or estimates made for the r
specific exposure period.
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5 Substitutingthederivedvalueofk into Equation 3.11-1 yields:
VC d-(0.057+k) T, 0.00087 (Eq. 3-9)
D r.
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o Table 3.11-1 Estimate of Exposure Rate at Critical Receptor From Vapor Container Shine.
EVC Made in Spring 1981 Monitoring Statio'n No.
d Direction k
k o
b VC (km)
(pR/hr)
(pR/hr)
(pR/hr) 13
.08 2250 20.4+
18.0*
.076 14
.11 3000 15.0 13.2*
.108 15
.08 3450 14.4 13.1*
.041 16
.13 300 17.4 16.6*
.067 17
.14 700 15.9 15.6 "
.029 18
.14 1150 24.9 24.6 "
.029 19
.16 1400 23.2 23.3 "
.013 20
.16 1600 19.4 18.8"
.076 21
.11 2050 20.3 18.7"
_._09 6 total 0.509
=
Average =
0.057 pR/hr All messures of E taker. 4/28/81, average daily power level was 97.5 MWe-Net. 5 days before shutdown at end of cycle 13/14.
Measures of 5 taken 5/13/81, during outage. Containment purged.
b Measures of k taken 6/2/81. during shutdown. Containment purged.
b i
i l
\\
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l C
N II 0
30 100 e
l',
d v.
c>
I,
,l
.o:* y
.y HETERS P
4. 16
/c 4*-
/
c-is 1 'is v..
.r l}
. f. '-
saa ex
}l s'
- . 1 -.... ;t s a gg,3 CM.14
---..i*eU.F Cate uu.eg ;
I I
- 5. t t c h.,e ogggee.
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- Ch.16 Turbine l
l '.
s Service Tt A 3 A $ r.;
@ gi,, j ) -.
Awa. sayl Area Cer.t e t
, - s yd Whee.
s, 4
s
, vapor e
,,...-p",,,~'y.-
g g
Containe r,e "j
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i f
s t
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L
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l i
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ee i
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Disposal I 1
e i,
l l
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a 6 cx-18 1
t i
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s b - g g;; yo-- - - - -.... -. @h.19
.i i
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Figure 4-4 Yankee Piant Radiologir.1 Environmental Monitoring Locations at the Restricted Area Fe*4ce (Direct Radiation Pathway)
/,/, /6.f Revision 6 - 2/16/68 4-5 Approved By:/ ',#g gnC..
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1 Table 4-1 Radiological Environmental Monitoring Stations
- Exposure Pathway Sample Location Distance From Direction From and/or Sample and Desianated Code +'
the Plant (km), the Plant 1.
AIRBORNE (Radiolodine and Particulate)
AP/CF-11 Observation Stand 0.5 NN AP/CF-12 Monroe Bridge 1.1 SW AP/CF-13 Rowe School 4.2 SE AP/CF-14 Harriman Power 3.2 N
Station AP/CF-21 Williamstown 22.2 W
2.
WATERBORNE a.
Surface WR-11 Bear Swamp Lower 6.3 Downriver Reservoir WR-21 Harriman Reservoir 10.1 Upriver b.
Ground WG-11 Plant Potable On-Site Well
.WG-12 Sharman Spring 0.2 NW c.
Sediment SE-11 Number 4 St: tion 36.2 Downriver From SE-21 Harriman Reservoir 10.1 Upriver Shoreline 3.
INGESTION a.
Milk TM-13 Whitingham, VT 8.4 ENE TM-12 Rendsboro, VT 6.1 N
TM-21 Williamstown, MA 21.0 WSW b.
Fish FH-11 Sherman Pond 1.5 At Discharge and Pet.it Inverter. FH-21 Harriman Reservoir 10.1 Upriver brates c.
Food TF-11 Monroe Bridge 1.3 SW Products TF-13 Monroe 1.9 WNW TF-21 Williamstown 21.0 WSW TV-11 Monroe Bridge **
1.3 SW l
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f Table 4-1 (continued)
Radiological Environmental Manitoring Stations
- Exposure Pathway Sample Location Distance From Direction From and/or Sample and Designated Code +
the Plant (km) the Plant 4
DIRECT RADIATION GM-1 Furlon House 0.8 SW GM-2 Observation Stand 0.5 NW GM-3 Rowe School 4.2 SE GM-4 Harriman Station 3.2 N
GM-5 Monroe Bridge 1.1 SW GM-6 Readsboro Road Barrier 1.3 N
GM-7 Whitingham Line 3.5 NE GM-8 Ponroe Hill Barrier 1.8 5
GM-9 Dunbar Brook 3.2 SW GM-10 Cross Road 3.5 E
GM-11 Adams High Line 2.1 WNW GM-12 Readsboro, VT 5.5 NNW GM-13 Restricted Area Fence 0.08 WSW GM-14 Restricted Area Fence 0.11 WNW GM-15 Restri ted Area Fence 0.08 NNW GM-16 Rest,cted Area Fence 0.13 NNE GM-17 Re1<.-lcted Area Fence 0.14 ENE GM-18 r es:ricted Area Fence 0.14 ESE GM-19 Restricted Area Fence 0.16 SE CM-20 Restricted Area Fence 0.16 SSE GM-21 Restricted Area Fence 0.11 SSW GM-22 Heartwellville 12.6 NNW GM-23 Williamstown Substation 22.2 W
GM-24 Harriman Dam 7.3 N
GM-25 Whitingham 7.1 NNE GM-26 Sadoga Road 7.6 NE GM-27 Number 9 Road 7.6 ENE GM-28 Number 9 Roao 6.0 E
GM-29 Route BA 8.2 ESE GM-3's Route BA 9.4 SE GM *,1 Legate Hill Road 7.6 SSE GM 32 Rowe Road 7.9 5
GM 33 Zoar Road 6.9 SSW GM-14 Fife Brook Road 6.4 SW GM-15 Whitcomb Summit 8.6 WSW GM-36 Tilda Road 6.6 W
GM-37 Turner Hill Road 6.7 WNW GM-38 West Hill Road 6.6 NW GM-39 Route 100 6.8 NNW GM-40 Readsboro Road 0.5 W
- Sample locations are shown on Figures 4-1 through 4-7.
- TV-11 Station is for leafy vegetables.
+ Station IX's are indicator stations and Station 2X's ate control stations (excluding the Direct Radiation stations).
~
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t 5.0 SETPOINT DETERMINATIONS Chapter 5,contains the basis for plant procedures that will meet the setpoint requirements.cf the Effluent Monitoring Instrumentation Technical Specifications (Specification 3.3.3.6 for ligulds and Sper.lfication 3.3.3.7 for gases).
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5.1 Liauid Effluent Instrumentation Setootnts Technical Specification 3.3.3.6 requires that the radioactive liquid effluent instrumentation in Table 3.3-8 of the Technical Specifications have alarm / trip setpoints in order to ensure Specification 3.11.1.1 is not exceeded.
That Specification limits the activity concentration in liquid effluents to the appropriate MPCs in 10CFR20, and a total noble gas MPC.
l Use the method below to determine the setpoints for the required instrumentation.
5.1.1 Method The instrument response (in counts per minute) for the limiting concentration at the point of discharge is the setpoint, denoted R, and is determined as follows:
f 3
(MPC ) (S )
(Eq. 5-1)
R. (f1 f) g g
2 where:
f)
. Flow rate past Test Tank monitor (gpm)
Flow rate past steam generator blowdown monitor (gpa) f2 Flow rate at point of discharge (gpm) f3 Instrument response factor (cpm /(pCl'/ml))
l Sg MPCc. Composite HPC for the mix of radionuclides (pCi/ml) fg/
f /MPC (Eq. 5-2)
Cg/
C /MPC MPC g
g-g g
c MPCt. MPC for radionuclide i from 10CFR20, Appendix B, Table 2, Column 2 (pCl/ml) t Cg
. Concentration of radionuclide i in mixture (pC1/ml)
Fraction of radionuclide 1 in mixture fg F
4!
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Other setpoint methodologies can also be applied which are more restrictive than the approach used here.
The setpoint, R, may be set lower to accommodate pathways without on-line monitors (secondary coolant or condensate leakage).
WhenHPCf15 not stable or when dilution flow 15 low, R may have to be ivaluated for each release.
5.1.2 Liquid Effluent Setpoint Example The effluent monitors for the Test Tank and steam generator blowdown i
release pathways are gamma sensitive monitors.
They both have a typical l
sensitivity, S, of 7.5E+7 cpm per 1 pC1/ml of gamma emitters which emit one photon per disintegration, and a typical background of 10,000 counts per minute.
Both monitors have adjustable alarm /setpoints.
However, the setpoint adjust control'is located inside the panel-mounted electronics cabinet and is not easily accessible.
The principal gamma emitting radionuclide in waste effluent streams is Xenon-133, averaging two orders of magnitude higher than any'other specie.
However, it is not the intent of effluent monitors to respond to dissolved noble gases, because Xenon-133 concentrations have never approached the MPC.
However, Iodine-131, Ceslum-134 and Cesium-137 are detected in every liquid effluent release in roughly equal quantitles and are the principal gamma emitters because they can approach their MPCs.
Therefore, for purposes of adjusting the alarm /setpoints of the effluent monitors to comply with 3.11.1.1, the composite MPC, HPC, of 6E-7 pC1/ml will be used.
g It is calculated based on the following data (to be conservative Iodine is weighted greater than the Ceslums):
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I f.
MPC g
g Cs-134
.5 3x10-7 fI l (Eq. 5-2)
MPC e
f jgpc 1
- 1/ (.25/9x10-6.25/2x10-5+.5/3x10-7)
+
6x10~7 g + f, is taken as 130 gpm, based l
The maximum liquid effluent flow rate, f 2
on a maximum 30 gpm flow rate from the Test Tank effluent pathway and a maximum 100 gpm flow rate from the steam generator blowdown pathway.
Both l
pathways will be assumed to operate continuously and simultaneously.
Dilution water flow, f, is taken as 140,000 gpm based on 138,000 gpm 3
through the condenstr and 2,000 gpm through the auxiliary cooling loop.
Throttling of cooling water is not practiced.
In this example, the setpoint for both monitors when both effluent pathways are operating is:
3 R
(f f) (MPC } (S )
(Eq. 5-1) c R
pCi/ml) (7.5 x 10*7 cpm /pC)/ml)
. (3) g m) (6.x 10-7 1
000 m
= 48,500 cpm Note that both effluent monitors have their lower level discriminators set to reject the pulses originating from the 80 kev gamma emissions from Xenon-133, and their high count rate alarms set at 48,500 cpm above background.
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e 5.1.3 Basis The 11guld effluent monitor setpoint must ensure tha,t Specification 3.11.1.1 is not exceeded for the appropriate in-plant pathways.
The monitor is placed upstream of the major source of dilution flow and responds to the concentration of radioactivity as follows:
I fsg)CMON (Eq. 5-5)
R = Sg(
g cpm i
where variables are the same as those in Section 5.1.1 except:
CHON = Total concentration (pC1/ml) seen by the monitor sg
. Ratio of response from equal activities of radionuclide i to a reference radionuclide Calibration of the radiation monitors have established that the gross gamma detectorresponse,5 Ifs.wasfairlyindependentofgammaenergy,as gg expected.
Thus,thehesponseisafunctionofradioactivityconcentrationand the gamma yield of the mixture.
Since fs is approximately one:
gg R - (S ) (Cggg)
(Eq. 5-6) g For simplicity, assume that both monitors look at the total flow for both, fg+f.
We know that:
2 f
+f C-(j p) (Cggg)
(Eq. 5-7) f 3 where:
C - Total concentration at point of discharge Solve Equation 5-5 for CMON and substitute into Equation 5-4 to get:
f 3 (C) (S )
(Eq. 5-8)
R-(f1,f) g 2
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We defined C =
C above and define MPC such that:
t g
g C
C, (Eq. 5-9)
MPC I
MPC c
1 1
The right side of the equation is the MPC limit in 10CFR20 solving for MPC, the composite MPC for the mixture, we get the definition of MPC :
g g
{C g
(Eq. 5-2)
=
g C g Substituting MPC into Equation 5-6, we get the response of the monitor as e
MPC is reached at the point of discharge, which is the setpoint:
g t
f i
3 (S )
(Eq. 5-1) l R = (f1 f ) (MPC )
g g
2 i
l I
t t
i i
t 1
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r 5.2 Gaseous Effluent Instrumentation Setpoints P
Technical Specification 3.3.3.7 requires that the radioactive gaseous effluent instrumentation in Table 3.3-9 of the Technical Spectftcations have their alarm setpoints set to ensure Specification 3.11.2.la is not exceeded.
That Specification limits the activity concentration in off-site gaseous effluents to well below the appropriate MPCs in 10CFR20 by limiting total body, skin and organ dose rate.
Use the method below to determine the setpoint for the noble gas activity monitor.
5.2.1 Method The noble gas activity monitor response (in counts per minute) at the limiting noble gas m -
(either total body or skin off-site) is the setpoint.
denoted R. and is determined as follows:
1 4
R 15 the lesser of:
1 (500) (60)
I
{ffsg)
(5 ) (
r 9
(Eq. 5-3)
Rtb "
N0 (F) (7.83) (
f DBF )
g g
i i
and ffs )
(3000) (60)
(5 ) (
g 9
(Eq. 5-4)
R5k.
ffDFj)
(F)
(
where:
l sg
. Ratto of response from equal activities of radionuclide i to a reference radionuclide, t.e., Xe-133 OFj
. Skin dose factor (see Table 1.1-2)
~
DF6g
. Total body dose factor (see Table 1.1-2) ff.Fractionactivityofradionuclideitototalnoblegasactivity Approved By: 4 d f
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- Primary vent stack flow rate (cc/ min)
S
. Instrument callbration factor (cpm /(pC)/cc))
g Other setpoint methodologies can also be applied which are more restrictive than the approach used here.
5.2.2 Gaseous Effluent Setpoint Example The primary vent stack noble gas activity monitor is an off-line system consisting of a beta sensitive scintillation detector, 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. Calibration data is provided by the manufacturer which indicates the r
response, s. of the beta sensit19e detector to various gaseous g
radionuclides.
The calibration data was verifted on installation and periodically thereafter.
System characteristics are:
Typical sensitivity I com = 3 x 10-8 pct /cc of Xenon-133; 7
that is, S = 3.3 x 10 cpm /(pC1/cc)
I' Typical background 10 to 20 cpm Under normal plant stack flow F, of 5.8x10 cc/ min (is 20,500 cfm x f
8 3
28,300 cc/ft ), one count on the scaler is equivalent to 17 microcuries of Xenon-133 noble gases released.
Since the typical average primary vent stack concentrations of noble gases are only about 1E-6 pCi/ce, direct grab sarpilng i
and isotopic analysis is not satisfactory.
The isotopic distributton of noble i
gases dissolved in primary coolant is determined monthly and used as the i
distribution,ff,forgaseouseffluentreleases.
The distribution.
I ff,andtherelativeresponse,s,foreachradionuclideinthis g
example are presented in Table 5.2-1.
Applying Equations 5-3 and 5-4:
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(3.3 x 10+7)
(0.71)
(500( (60)
(2.1 x 10-3) - 73,700 cpm Rtb -
(5.8 x 10+8) (7.83)
(3.3 x 10*I) (0.71)
(3000) (60) - 169,000 epm R5k "
8 (4.3 x 10-2)
(5.8 x 10 )
The setpoint, R, is the lesser of R and R and is, therefore, tb sk 73,700 cpm.
This is because for the noble gas mixture in this example, the l
total body dose rate is more restrictive.
(
t 5.2.3 Basis The noble gas activity monitor setpoint must ensure that Specification 3.11.2.la is not exceeded.
Sections 3.4 and 3.5 show that Equations 3-3 and 3-4 are acceptable methods for complying with that Specification.
Which equation (i.e., dose - total body or skin) is more limiting depends on the noble gas mixture.
Therefore, each equation must be considered separately.
The derivation of Equations 5-3 and 5-4 starts with the general equation for the response. R, of a radiation monitor (in epm):
ffs)
(C)
(Eq. 5-5) 9
.(
R
. (5 )
g (cpm)
(cpm /(pC1/cc)) (1) (pCi/cc) or, expanding for the concentration:
e ffs)
(6) (60 / F)
(Eq. 5-10)
(5 ) (
R g
9 (com)
(cpm /(pC1/cc))(1)(pCl/sec)(sec/ min)(cc/ min)
The response of the monitor at the release rate which causes the total body dose ratt Ilmit to be reduced, Rtb, begins with Equation 3-3:
btt " I'03 fI DFB 3 substnetieg6.fp6,gives:
uq. 5-in Revision 6 - 2/18/88 Approved By: A[
5-9
/
!004R/12.243
p g'
I btb = 7.83 h [l ff0FB, (Eq. 5-12) rearranging to solve for Q:
h=
tb (Eq. 5-13) 7.83Iff0FB3 1
Substituting Eq. 5-13 into Eq. 5-10 and substituting the total body dose rat' ilmit gives:
S(cpm /(pC1/cc))( Iffs)500(mrem /yr) 60(sec/ min) g (Eq. 5-3)
Rtb
- F'(cc/ min) 7.83 (pci-sec/pCl-m )
.ffG0FB (mrem-m /pCl-yr) 3 3
g The response of the monitor at the release rate which causes the skin dose rate limit to be reduced. Rg, begins with Equation 3-4:
b,g-
{hgDF'g substitutingh=ffhg gives:
(Eq. 5-11) 3g=h{ff0Fj b
(Eq. 5-14)
Rearrangingtosolveforh:
b h=
5 (Eq. 5-15) i OIb Substituting Eq. 5-15 into Eq. 5-10 and substituting the skin dose rate limit of 3,000 mrem /yr gives:
Revislon 6 - 2/18/88 4 proved By:/#,
~
5-10 5004R/12.243
=
4
'O S (cpm /(pCi/cc))(
ff0s ) 3000(mrem /yr) 60(sec/ min) g g
Rsk
- ff0Fj(mrem-sec/pCl-yr)
F(cc/ min) 5004R/12.243
?b
~
~
Table 5.2-1 Sample Calculation of Gaseous Instrumentation Setpoint (Based on 1981 Yankee Data)
Detector
Response
Weighted Weighed Fraction With Weighted Whole Body Skin Dose Noble Gas of Totti Respect to
Response
Dose Factor Factor NG f 'C x DFB f
xDFj ff0 Specie ff Xe-133 1.0 s xs g g
g g
Ar-41 0.008 1.2 0.01 7.1E-5' 1.2E-3 l
Kr-85 0.000 1.15 0
-0
-0 Kr-85m 0.010 0
0 1.2E-5 4.6E-4 Kr-87 0.010 1.5 0.015 5.9E-5 2.9E-3 Kr-88 0.016 1.15 0.018 2.3E.4 3.0E-3 Xe-131m 0.020 0
0 1.8E-6 2.5E-4 Xe-133 0.38 1.0 0.38 1.1E-4 4.0E-3 Xe-133m 0.000 0
0
-0
-0 Xe-135 0.20 1.3 0.26 3.6E-4 1.2E-2 Xe-135m 0.34 0
0 1.1E-3 1.6E-2 Xe-138 0.02 1.5 0.03 1.8E-4 3.6E-3 Summation 1.0 0.71 2.1E-3 4.3E-2 ApprovedBy:/h[
Revision 6 - 2/18/88 5-12
[
5004R/12.241
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- o Table 5.2-2 Relative Fractions of Core Inventory i
Noble Gases After Shutdawn Time Kr-85m Kr-85
<r-88 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 l
t < 24 H
.007
.003
.004
.004
.021
.714
.017
.232
.005
.023
.911
.001
.056
.004 24 hr i t < 48 h
.005
.008
.015
.971 48 h I t < 5 d l
.010
.013
.006
.970 5 d I t < 10 d
.022
.002
.956 10 d I t < 15 d
.020 i
.034
.001
.929
.037 l
15 1 t < 20 d 1
.071
.806 l
.119 20 1 t < 30 d
.103
.103
.795 30 1 t < 60 d I
.974
.024
.002 t1 60 d f
f l
Revision 6 - 2/18/88 Approved By: A>b f
5-13 5004R/12.243 I
I l-
(
- ).
this header.
The water from this drain header discharges without further dilution into a tributary of the Deerfield River outside the controlled area.
A composite sampler collects a sample of the water whenever there is discharge (water in the pipe).
Batch effluent tanks called "test tanks" collect the distillate from the liquid radioactive waste evaporator, Normally, liquid waste accumulates at about I gpm and is processed at about 4 gpm. When a 7000 gallon test tank is filled, it is sampled, analyzed, and released at a nominal 30 gpm.
The condenser cooling flow provides the major source of dilution and is assumed to be 138,000 gpm with two pumps operating and 69,000 gpm with one pump operating.
Throttling of condenser cooling water is not practiced at Yankee Plant.
During shutdown periods, the 4,000 gpm service water provides dilution Wdter flow.
Flow rate is Varlable and estimated by pump curves.
Typically flow rates range from 1,500 to 3,500 gpm.
The discharge rate from the steam generator blowdown tank is fixed by piping geometry and a relatively constant head on the tank. A flow meter estimates the discharge rate during periods of discharge.
Verification is done periodically by measuring the time for the tank level to decrease during a normal release.
The discharge rate for the Turbine Building pathway is estimated to be 400 gpm. Appt timately I gpm of this is secondary coolant (from pump leakage and sample stations) the remainder is service water from various secondary plant heat exchangers. All piping is buried and inaccessible so flow is estimated from cooling water pump flows.
The discharge rate for the test tanks is controlled by the discharge line vart-ortftces and limited to 30 gpm.
Calibration of the radiation monitors have established that the gross gamma detector response was fairly independent of the gamma energy, as expected.
Thus, the response is a function of the radioactivity y/
Revision 6 - 2/18/88 Approved By: Af MM 6-2
4 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 1", U. S. Nuclear Regulatory Commission, Revision 1, October 1977.
B.
Hamawl, 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 1977.
C.
Regulatory Guide 1.111. "Methods for Estimating Atmospheric Transport and D1spersion of Gaseous Effluents in Routine Releases From Light-Water Cooled Reactors " U. S Nuclear Regulatory Commission, March 1976.
D.
NEP 1 and 2 Preliminary Safety Analysts Report, New England Power Company, Dockat Nos. STN 50-568 and STN 50-569.
E.
Yankee Atomic Technical Specifications.
F.
Yankee Atomic Electric Company Supplemental Information for the Purposes of Evaluation of 10CFR50, Appendix I, Amendment 2, October 1976.
(Transmitted by J. L. French - YAEC to USNRC in letters dated June 2, 1976 August 31, 1976, and October 8, 1976.)
G.
National Bureau of Standards, "Maximum Permissible Body Burdens and Maximum Permissible Concentrations of Radionuclides in Air and in Water for 0:cupational Exposure," Handbook 69. June 5, 1959.
H.
Slade, D. H., "Meteorology and Atomic Energy - 1968," USAEC, July 1968.
_A e[
f Revision 6 - 2/18/88 Approved By:
R-1 5046R/26.216
it APPENDIX X-Radioactive Liquid. Caseous, and Solid Waste Treatment Systems Requirement: Technical Specification 6.16.1 requires that licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) be reported to the Commission in the Semiannual Radioactive Ef fluent Release Report for the period in which the evaluation was reviewed by the Plant Operation Review Committee.
i Responset There were no licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) during this reporting period.
e i
i
[
i i
t I
i l
t L
I f
4824R/20.124
y
- L,
APPENDIX !
l Supplemental Information F*rst and Second Quarters. 1988 1.
' Technical Specification Limits - Dose and Dose Rate Technical Specifica ton and Cateaory Limit a.
Noble Gases 3.11.2.1 Total body dose rate 500 ares /yr 3.11.2.1 Skin dose rate 3000 ares /yr 3.11.2.2 Gamma air dose 5 mrad in a quarter 3.11.2.2 Gamma air dose 10 mrad in a year 3.11.2.2 Beta air. dose 10 mrad in a quarter 3.11.2.2 Beta air dose 20 mrad in a year b.
Iodine-131. Tritium an'd Radionuelides in Particulate Form k'ith Half-lives Creater than 8 days 3.11.2.1 Organ dose rate 1500 arem/yr 3.11.2.3 organ dose 7.5 arem in a quarter 3.11.2.3 organ dose 15 area in a year c.
Liquids 3.11.1.2 Total bcdy dose 1.5 area in a quarter 3.11.1.2 Total body dose 3 area ir a year 3.11.1.2 Organ dose 5 arem in a quarter 3.11.1.2 organ dose 10 area in a year I-1 4824R/20.124
L a
APPENDIX !
(Continued) 2.
Technical Specification Limits - Concentration Technical Specification and Category Limit a.
Noble Cases No MPC limits 3
b.
Iodine-131. Tritium and Radionuelides No MPC limits in Particulate Form With Half-Lives Creater than 8 days c.
Liquids 3.11.1.1 Total sum of the fraction of MPC (10CTR20, Appendix B, Tables.II, Column 2), excludirs nobis gases less than:
1.0 3.11.5. 1 Total noble gas concentration 2E-04 uC1/cc l
3.
Measurements and Approximations of Total Radioactivity a.
Noble Cases "Continuous discharges" are determined by indirect measurement.
Primary gas samples are taken periodically and analyzed.
It le assumed *. hat in primary to secondary leakage all gases are ejected through the air ejector.
In primary coolant charging pump leakage all gases are ejected to the primary vent stack either during flashing or liquid waste processir.g.
"Batch discharges" are determined by direct measurement. Errors associated with these measurements are estimated to be 155 percent.
I-2 4824R/20.124
O o
s APPENDIX !
(Continued) b.
Iodines todines are continuously monitored by drawing a sample from the primary vent stack through a particulate filter and char:oal cartridge. The filter and charcoal cartridge are removed and analysed weekly. The errors associated with these measurements are estimated to be 125 porcent.
c.
Particulatis The part.culate filter described in (b) above is analyzed weekly.
The errors associated with the determination of particulate effluents are estimated to be 130 percent.
d.
Liquid Effluent,s Liquid effluents are determined by direct measurement.
In line composite samples are analyzed for strontium - 89, strontium - 90, iron - 55. gross alpha activity and carbon - 14 There is no compositing of samples for tritium or dissolved fission gas analysis. For continuous discharges composite samples are used for gamma isotopic analysis. A ganma isotopic analysis is performed on a representative sample for each batch release using the Marinelli Beaker geometry. The errors associated with these measurements are as followst fission and activation products, 120 percenti tritium, 210 percenti dissolved fission gases. 120 percenti alpha activity, 135 percoat.
(
l l
I-3 4824R/20.124 iE
y
.aw e
%b APPENDIX I (Continaed) u, 4.
Batch Releases a.
Liquids First Quarter i
. Number of batch releases:
11 Tota' time period for batch releases: 4321 minutes
' ni.
- ime period for a batch release:
1815 minutes ge time per.tod for batch releases:
393 minutes Minimum time period for a batch' release:
165 minutes Aversge stream flow during period (Sherman Dam):
748 cfs Average discharge rate:
15.0 gpm Second Quarter Number of batch releases:
18 Total time period for batch releases:
7586 minutes Maximum time period for a batch release:
1824 minutes Average time period for batch releases:
237 u*nutes Minimum time period for a batch release 421 minuter Average streem flow during period (Sherman Dam): 442 cfs Average discharge rate:
18.0 gpm b.
Cases l
There were no batch releases during the firrt and second quarters.
I-4 4824R/20.124
,, e fy:.,y /,,., 1 ;
/..-
APPENDIX I
'(Continued) 4 5.
Abnormal Releases 4
a.
Liquid-4 There were no nonroutine liquid releases during the reporting period.
b.
Gasej There were no nonroutine gaseous releases during the report J.ig period.
d
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