ML20035C977
| ML20035C977 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Perry, Byron, Braidwood, Quad Cities, Zion, LaSalle  |
| Issue date: | 02/26/1993 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| References | |
| PROC-930226, NUDOCS 9304120006 | |
| Download: ML20035C977 (100) | |
Text
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A_bulhluhY COMMISs10h n1L si,alun el-13/
February 26, 1993 i
to itJ h 1 I, b i d h bb 2cb >
Attached are the recent revisions to the Offsite Dose Calculation Manual's (ODCM) Generic Section for all six stations.
Please complete the following manual update.
1 REMOVE D11ERJ Generic ODCM Revision Index Generic ODCM Revision index Revision 0.A Revision 0.K
- p. i to xii
- p. i to xii Table of Contents Table of Contents Revision 0 Revision 0.K, Jan. 1993 p. 14
- p. 14 j
Chapter 3 Chapter 3 f
Revision 0.A, April 1991 Revision 0.K, Jan. 1993
- p. 5, 6
- p. 5, 6 Chapter 4 Chapter 4 Revision 0.K
- p. 25, 27, 28, 31, 32
- p. 25, 27, 28, 31, 32 Chapter 7 Chapter 7 Revision 0.A, April 1991 Revision 0.K, Jan. 1993
- p. 3
- p. 3 Chapter 8 Chapter 8 Revision 0.A, April 1991 Revision 0.K, Jan. 1993
- p. 2, 3
- p. 2, 3 Reference Section Reference Section Revision 0.A, April 1991 Revision 0.K, Jan. 1993 p.5,6,7,8 p.5,6,7,8 1
I I
l 060062 t
I ZEMERPLN/347/27 9304120006 930226 DR ADOCK 05000010 p
]
A..
February 26, 1993 (Cont'd)
REMOVE INSERT Appendix A Appendix A Revision 0.K, Jan. 1993 pgs. 2 - 5, 7 - 10, 12 - 14, pgs. 2 - 5, 7 - 10, 12 - 14, pgs. 16 - 20, 23, 25 - 39, 41, pgs. 16 - 20, 23, 25 - 39, 41, pgs. 42, 44, 45, 47, 49 - 54, pgs. 42, 44, 45, 45a, 47, 49, pgs. 56, 58 - 60 pgs. 50 - 54, 54a, 56, 58 - 60 Appendix D Appendix D Revision 0.K, Jan. 1993 pgs. 1, 17, 18, 19, 20, 32 pgs. 1, 4a, 4b, 4c, 13a, 13b, pgs. 13c, 17, 18, 19, 20, 32 Appendix E Appendix E Revision 0.K Jan. 1993
- p. 8, 26, 29, 30 pgs. 8, Ba, 26, 29, 30 f
Please sign and date this contiol sheet and return to:
COMMONWEALTH EDISON COMPANY c/o Document Control - Emergency Preparedness 1400 Opus Pl.- 5th Floor Downers Grove, IL 60515 (708)663-6547 Your signature indicates you have verified that your control number is f
correct and you have updated your manual.
l Signature Date i
1 ZEMERPLN/347/28
REVISION 0.K JANUARY 1993-l GENERIC ODCM REVISION INDEX PAGE REVISION O
GENERIC ODCM REVISION INDElt j
i 0.K ii 0.K i
iii 0.K iv 0.K v
0.K vi 0.K vii 0.K viii 0.K ix 0.K x
0.K xi 0.K xii 0.K 1
i
SUMMARY
OF TOPICS i
TOC-1 0.A I
GFEERIC TABLE OF CONTENTS TOC-2 0.A l
TOC-3 0.A j
TOC-4 0.A l
TOC-5 0.A GEEERIC LIST OF FIGURES TOC-6 0.A GENERIC LIST OF TABLES TOC-7 0.A GENERIC LIST OF APPENDICES l
TOC-8 0.A TOC-9 0.A j
s TOC-10 0.A TOC-ll 0.A i
TOC-12 0.A TOC-13 0.A TOC-14 0.K TOC-15 0.A TOC-16 0.A LIST OF ABBREVIATIONS AND ACRONYMS TOC-17 0.A O
TOC-18 0.A TOC-19 0.A TOC-20 0.A l
l 1
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION O
ammu 1-i 0.A 1-ii 0
1-iii 0.A 1-1 0
1-2 0.A 1-3 0.A 1-4 0.A 1-5 0.A 1-Sa 0.A 1-6 0.A 1-7 0
1-8 0
1-9 0.A 1-10 0
1-11 0.A 1-12 0.A 1-13 0.A O
O ii l
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION CHAElIR_2 2-i 0.A 2-ii 0.A 2-1 0
2-2 0
2-3 0.A 2-4 0.A 2-4a 0.A 2-5 0.A 2-6 0.A 2-7 0.A 2-8 0.A 2-Ba 0.A 2-8b 0.A 2-8c 0.A 2-9 0.A 2-10 0.A 2-11 0.A 2-12 0.A 2-13 0.A i
2-14 0
l 2-15 0
2-16 0.A 2-17 0.A CHAPTER 3 3-i 0
3-ii 0.A 3-1 0.A i
3-2 0.A 3-3 0.A 3-4 0
3-5 0.K 3-6 0.K (3
L) iii
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEI (Cont'd)
PAGE REVISION CIIAPTER 4 4-i 0
4-ii 0.A 4-iii 0
4-1 0.A l
4-2 0.A l
4-3 0.A 4-4 0
4-5 0.A 4-6 0.A 4-7 0
4-8 0
4-9 0.A 4-10 0
4-11 0.A 4-12 0.A 4-12a 0.A 4-13 0
l 4-14 0.A 4-15 0.A 4-16 0.A 4-17 0
4-18 0
[~'
4-19 0
\\
4-20 0
4-21 0
4-22 0
4-23 0
4-24 0
i 4-25 0.K 4-26 0
4-27 0.K 4-28 0.K 4-29 0
4-30 0
4-31 0.K 4-32 0.K 4-33 0.A 4-34 0.A 4-35 0
4-36 0.A 4-37 0
4-38 0
V iv
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION CIMETER__5 i
5-i 0
t 5-1 0.A l
5-2 0.A
(
5-3 0
5-4 0.A l
5-5 0.A l
5-6 0.A 5-7 0.A i
CtMPTER_6 l
6-i o
6-1 0.A 6-2 0.A 6-3 0.A 6-4 0.A 6-5 0.A I
CliAEIZB 7 7-i 0
7-ii 0
(
7-iii 0.A L
7-1 0.A l
7-2 0.A 7-3 0.K 7-4 0.A 7-5 0.A 7-6 0
7-7 0.A 7-8 0.A 7-9 0.A 7-10 0.A 7-11 0
7-12 0
7-13 0
7-14 0
7-15 0
7-16 0
7-17 0
7-18 0
l l
V
l REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION C MPTER 8 l
8-i 0
8-1 0.A 8-2 0.K 8-3 0.K CHAPTER 9 9-1 0.A REFEREN_CES j
R-1 0.A R-2 0.A R-3 0.A i
R-4 0.A R-5 0.K R-6 0.K R-7 0.K R-8 0.K w
q%J vi
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION hEPLK121X_b A-i 0.A A-ii 0.A A-iii 0.A A-iv 0
A-1 0.A A-2 0.K A-3 0.K A-4 0.K A-5 0.K A-6 0.A A-7 0.K A-8 0.K A-9 0.K A-10 0.K j
A-ll 0.A A-12 0.K l
A-13 0.K A-14 0.K A-15 0.A A-16 0.K A-17 0.K A-18 0.K O-A-19 0.K A-20 0.K A-21 0.A A-22 0.A A-23 0.K A-24 0.A A-25 0.K A-26 0.K A-27 0.K A-28 0.K 1
A-29 0.K A-30 0.K A-31 0.K A-32 0.K A-33 0.K l
A-34 0.K l
A-35 0.K l
A-36 0.K A-37 0.K A-38 0.K A-39 0.K A-40 0.A A-41 0.K A-42 0.K A-43 0.A
' (]/
(~
A-44 0.K l
A-45 0.K l
vii I
l
REVISION 0.K l
JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd) f PAGE REVISION I
r r
! k APPENDIX A_LCo_nt'd)
A-46 0.A A-47 0.K A-48 0
A-49 0.K i
i A-50 0.K A-51 0.K
]
A-52 0.K A-53 0.K A-54 0.K A-54a 0.K A-55 0.A l
l A-56 0.K l
A-57 0.A 1
A-58 0.K i
A-59 0.K l
A-60 0.K l
A-61 0
O O
Viii
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION
/]
/
APPENDIX B B-i 0
E-ii 0.A B-iii 0
B-iv 0
B-1 0.A B-2 0
B-3 0.A B-4 0.A B-5 0
i B-6 0
B-7 0
B-8 0
B-9 0
B-10 0
B-ll 0
B-12 0.A B-13 0
B-14 0.A B-15 0
B-16 0
B-17 0.A B-18 0.A 7.
l B-19 0
k B-20 0.A B-21 0.A B-22 0.A
~
j B-23 0.A B-24 0
B-25 0.A B-26 0
B-27 0.A B-28 0.A B-29 0
B-30 0.A B-31 0.A B-32 0.A l
B-33 0.A B-33a 0.A B-34 0
B-35 0
B-36 0
B-37 0
B-38 0
\\~
ix
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION APPENDIX C l
C-i 0
C-ii 0
C-1 0.A C-2 0.A C-3 0.A C-4 0.A C-5 0.A i
C-6 0.A C-7 0
C-8 0
C-9 0
l l
i
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION APPENDIX D D-i 0
D-ii 0.A D-iii 0
D-1 0.K D-2 0
D-3 0
D-4 0
D-4a 0.K D-4b 0.K D-4c 0.K D-5 0
D-6 0
D-7 0
D-8 0
D-9 0
D-10 0
D-11 0
D-12 0
D-13 0
D-13a 0.K D-13b 0.K D-13c 0.K D-14 0
O D-15 0
D-16 0
D-17 0.K i
D-18 0.K D-19 0.K D-20 0.K 1
D-21 0
D-22 0
l D-23 0
D-24 0.A D-25 0.A D-26 0.A D-27 0
D-28 0
D-29 0
D-30 0.A D-31 0.A D-32 0.K D-33 0
D-34 0
D-35 0
D-36 0
(
D-37 0
D-38 0.A D-39 0
(N
/
D-40 0
)
D-41 0
D-42 0
D-43 0
l D-44 0
j D-45 0
D-46 0
xi l
REVISION 0.K JANUARY 1993 GENERIC ODCM REVISION INDEX (Cont'd)
PAGE REVISION O
APPENDIX _E E-i 0.A E-ii 0
E-iii 0
E-iv 0.A E-1 0.A E-2 0
E-3 0.A E-4 0.A E-5 0.A E-6 0.A E-7 0.A E-8 0.K E-8a 0.K E-9 0.A E-10 0.A E-11 0.A E-12 0.A E-13 0A E-14 0.A E-15 0.A E-16 0.A E-17 0.A C
E-18 0
(
E-19 0
E-20 0
E-21 0
E-22 0.A E-23 0
E-24 0
E-25 0.A E-26 0.K E-27 0.A E-28 0.A E-29 0.K E-30 0.K E-31 0.A E-32 0.A E-33 0
E-34 0.A E-35 0.A E-36 0.A E-37 0.A E-38 0
E-39 0
E-40 0
E-41 0.A E-42 0
(]
E-43 0
(,/
E-44 0
E-45 0
E-46 0
E-47 0
xii
REVISION 0.K JANUARY 1993 GENERIC LIST OF TABLES - APPENDICES NUMBER TITLE PAGE A-1 Release Point Classifications A-58 A-2 Nearest Downstream Community A-59 Water Systems Affected B-1 Portion of an Example Joint B-34 Frequency Distribution C-1 Illustration of Model for Calculating C-8 Dose Due to Radioactivity Releases C-2 Illustration of Model for Dilution of C-9 Tank Discharges D-1 Inhalation Dose Factors for Adults D-2 D-la Inhalation Dose Factors for Teenager D-4a l
D-2 Inhalation Dose Factors for Children D-5 D-3 Inhalation Dose Factors for Infants D-8 D-4 Ingestion Dose Factors for Adults D-ll D-4a Ingestion Dose Factors for Teenager D-13a {
D-5 Ingestion Dose Factors for Infants D-14 D-6 Miscellaneous Dose Assessment Factors D-17 Environmental Parameters D-7 Miscellaneous Dose Assessment Factors D-19 Consumption Rate Parameters D-8 Stable Element Transfer Data D-21
/"
D-9 Atmospheric Stability Classes D-23
'x D-10 Vertical Dispersion Parameters D-25 D-ll Maximum Permissible Concentrations (MPCs) of D-27 Dissolved or Entrained Noble Gases Released from the Site to Unrestricted Areas in Liquid Waste D-12 Radiological Decay Constants (A )
D-28 i
D-13 Bioaccumulation Factors Bi to be Used in the D-32 Absence of Site-Specific Data D-14 Beta Air and Skin Dose Factors for Noble Gases D-34 D-15 External Dose Factors for Standing on D-35 1
Contaminated Ground D-16 Sector Code Definitions D-38 1
l'
(.
TOC-14
REVISION O.K JANUARY 1993
]
O
-Inhalation i
Milk Ingestion +
i ingestion +
Animals Most ingestion +
Air ->
-> Man j
~
- Deposition -D Vegetation -ingestion i
j I
- Do po sition.-> Soll
- Radiation 5
- Radiation i
l O
All of these pathways are considered in Regulatory Guide 1.109 (Reference 6).
In the station's calculations, only the solid-line pathways are considered.
The dashed-line pathway (uptake of radioactivity by vegetation from soil) is omitted.
l 3-5
REVISION O.K JANUARY 1993 O
Potable
> water Ingestion y
supply 8 8
-Uptake A
--In g e stio n i
tooda Milk ingestion -+
--I n g e s t i o n -------------- + Animals
,-Ing e stion.5 I
Meat Ingestion -->
s 8
+ Man Water +
- -Irrig a tio n --> Vegetation [ -I n g e s t i o n - - - - - - - - - - - - - - - - - - - - - - - -+
4 1
-- Irrig a tio n --@ soll
-- De po sition --.
Shoreline
-- R a d i a t io n - - ------ ----- - ------------ -- #
sediment
((
All of these pathways are considered in Regulatory Guide 1.109 (Reference 6).
In the station's calculations, l
only the solid-line pathways (ingesting of potable water and aquatic foods) are considered.
O 3-6
REVISION O.K JANUARY 1993 D nhalja Inhalation Dose Commitment
[ mrem]
i O
Dose commitment to organ j of an individual in age group a due to inhalation of non-noble-gas radionuclides released in gaseous effluents.
I ia Inhaled Radioactivity
[pCi]
Activity of radionuclide i inhaled by an individual in age group a in the time period under consideration.
DFA ja Inhalation Dose Factor
[ mrem /pCi]
i Dose commitment to organ j of an individual in age group a per unit of activity of radio-nuclide i inhaled.
The amount of radioactivity inhaled Iia [pCi] is the product of 3
radioactivity concentration in air [pCi/m ],
breathing rate R a 3
[m /yr], and the time period of concern [yr].
The radioac-tivity concentration is the product of radioactivity release 3
rate [pCi/sec] and relative concentration factor X/Q [sec/m ),
The Inhalation Dose Commitment Factor Values of the inhalation dose commitment factor are the same for all stations.
This is because the factor is defined as a dose rate per unit of radioactivity intake.
The station-
~
specific aspects of the calculation of inhalation dose concern the quantity of radioactivity inhaled.
Values of the inhalation dose commitment factor are provided in s
Tables D-1, D-la, D-2, and D-3 of Appendix D of this manual as follows:
For various potential effluent non-noble-gas radio-nuclides.
e For four age groups:
Adult (17 years and older).
Teenager (11 years to 17 years).
Child (1 to 11 years).
4-25 I
REVISION O.K JANUARY 1993 1
e Leafy vegetables.
O e
Produce (nonleafy vegetables, fruit, and grain).
I e
Milk.
e Meat.
Equation A-18 of Appendix A is used to calculate the dose commitment due to ingestion of food containing non-noble-gas radionuclides released in gaseous effluents.
The dose commit-ment due to ingestion of a-single radionuclide i may be represented as follows:
f00d D
ja
- Iia DFI ja (4-11) i l
f00dja-Food Pathways Dose Commitment
[ mrem]
D Dose commitment to organ j of an individual in age group a due to l
ingestion via food pathways of non-noble-gas radionuclides released in gaseous effluents.
Iia Ingested Activity
[pCi]
Activity of radionuclide i ingested by an individual in age group a in the time period under consideration.
DFI ja Ingestion Dose Commitment Factor
[ mrem /pCi]
i Dose commitment to organ j of an individual in age group a per unit of activity of radionuclide i ingested.
Values of the ingestion dose commitment factor are the same for all stations.
This is because the1 factor is defined as a dose j
rate per unit of radioactivity intake.
The station-specific aspects of the calculation of ingestion dose concern the quan-tity of radioactivity ingested.
Values of the ingestion dose commitment factor are provided in Tables D-4, D-4a, D-4b and D-5 of Appendix D of this manual for various radionuclides, for O
l l
4-27 l
REVISION O.K JANUARY 1993 members of the public, and for the same organs as the O
inhalation dose commitment factor is provided for.
The ingested activity Iia is calculated by an equation of the following form (see Equations A-19 through A-22 of Appendix A):
Iia = Ci Ua (tr/365) f (4-12)
Ci Food Product Radioactivity
[pCi/L or pCi/kg]
Concentration Average concentration of radionuclide i in the food product during the time period of interest.
The units are pCi/L for milk and pCi/kg for leafy vegetables, produce, and meat.
U Food Product Consumption Rate [L/yr or kg/yr]
a Annual consumption (usage) rate of the food product for individuals in age group a.
The units are L/yr for milk and kg/yr for leafy vegetables, produce, and meat.
t Time Period of Interest
[ days]
r Time period of release or exposure.
1/365 Conversion Constant
[yr/ day]
Converts days to years.
f Food Product Affected Fraction Fraction of the consumed food product that is affected by radioactivity released from the plant.
For milk and meat', f is taken as 1.
For leafy vegetables, it is taken as fy, and for produce it is taken as f p.
The food product radioactivity concentration is calculated from measurements of radioactivity in station releases.
The differ-ent equations used for radioactivity concentration in vegeta-tion, milk, and meat are discussed balow.
O 4-28
i REVISION O.K I
JANUARY 1993 l
e The average concentration C i in feed [pci/kg].
f The animal's feed consumption rate Wg [kg/ day].
e i
The. fraction Fy of an animal's daily intake of radio-e activity that appears in meat (pCi/kg in meat per pCi/ day ingested by the animal) (days /kg].
e A factor to account for radiodecay between the slaugh-ter of the animal and the consumption of its meat.
j 4.3 LIQUID RELEASES The evaluation of doses and dose rates due to radioactivity in liquid effluents is required to assess compliance with provi-sions of RETS related to 10 CFR 50, Appendix I.
Some of the radioactivity released in liquid effluents from a nuclear power station may be ingested by persons who drink I
water or consume fish from a body of water receiving station liquid discharges.
The stations obtain the dose commitment due
'(
to radioactivity in liquid releases as the sum of dose commitments from the drinking water and fish pathways.
The dose commitment is calculated for an adult.
Equations A-30 and A-31 of Appendix A are used to calculate dose commitments for the members of the public due to consumption of drinking water and fish.
The total dose commitment for the liquid pathway due to a radionuclide i is as l
follows:
s Dii9 f
DFI ja + I ia DFI ja (4-13) ja " I ia i
i 1i9 Dose Commitmbnt for a Member of the
[ mrem]
D 3
public Due to Radioactivity in Liquid Effluents O
1 P
4-31 l
REVISION O.K JANUARY 1993 O
Dose commitment to organ j of a member of the public of age group a consuming drinking water and fish containing radioactivity released in liquid effluents.
W I ias I ia Ingested Activity
[pCi]
Activity of radionuclide i ingested by an individual in age group a in the time period under consideration due to consugption of W
drinking water (I ia) or fish (I ia)-
DFl ja Ingestion Dose Factor
[ mrem /pCi]
i Dose commitment to organ j of an individual in age group a per unit of activity of radio-nuclide i ingested.
The ingested activity is calculated as follows:
W W
W I ia = U C i t (4-14) f f
f I ia = U C i t (4-15)
W f
U, U Usage Factor
[L/hr, kg/hr]
a a
W f
Consumption rate of water (U ) or fish (U )
by individual in age group a.
W f
C i, C i Concentration of
[pCi/L, pCi/kg)
Radioactivity Concentration of radioguclide i in drinking water (CW-or fish (C i) due to release in liquid efb)uents.
l t
Time Period of Concern
[hr]
Time period during which the relea'se and consumption occur.
The radioactivity concentration in water CW i [pCi/L] is obtained by dividing the radioactivity release [pCi] by the volume of water [L] in which the release is diluted (e.g.,
the 4-32
REVISION O.K JANUARY 1993 Stations entitle this report, " Annual Radiological Environmental-j
()
Operating Report
" where is the year covered in the report.
All stations except Dresden must submit it prior to May 1 of each year; Dresden must submit it by March 31.
The report contains results for the preceding calendar year.
i There are slight variations from station to station in the j
required content of this report (see the station RETS).
However, items required at one or more station (s) include the following-t e
An assessment of the radiation doses from radioactive liquid and gaseous effluents during the previous calendar year.
i An annual summary of hourly meteorological data e
collected over the previous year, which may be on I
magnetic tape.
e An assessment of radiation doses from the station and l
nearby uranium fuel cycle sources'to the likely most exposed member of the public.
The purpose is to show i
O conformance with 40 CFR 190.
i A summary description of the radiological environmental e
monitoring program.
e Results of the analyses of all radiological environ-mental samples and of all environmental radia-tion measurements during the report period.
e A summary and tabulation of the above results in a particular format.
The stations use a format like that shown in Figure 7-1.
This format involves use of the indicator / control concept (see Section 5.3).
For each sample type and analysis, the summary table incledes one column which contains the cean and range of results at indicator locations and another column which contains the mean and range of results at control locations.
e Summaries, interpretations, and an analysis of trends of_the results of the radiological environmental surveillance activities for the report period.
e Results of the annual land use census.
O 7-3
REVISION O.K JANUARY 1993 i
implements the PCP for solid radwaste.
It measures hydrogen or oxygen concentration in off-gas.
It also measures tank i
radioactivity and BNR off-gas radioactivity.
The station maintains instrumentation associated with these activities and demonstrates operability of the instrumentation in accordance with the surveillance requirements of the RETS.
In the event that any RETS requirements are violated, the station is responsible for taking one of the actions allowed by 4
the RETS and issuing any required reports to the NRC.
j i
The station assembles the Semiannual Radioactive Effluent j
Release Report, issues the report, and distributes it in accordance with Reference 42.
The station also issues the Annual Radiological Environmental Operating Report.
The report is assembled by the environmental contractor.
NSEP reviews the report prior to issuance and
(
distributes the report after it is issued.
8.2 METEOROLOGICAL CONTRACTOR The meteorological contractor operates and maintains the meteorological tower instrumentation at each station.
The contractor collects and analyzes the data and issues periodic reports (weekly, monthly and semiannually) to NSEp.
The contractor prepares the meteorological data summary required I
for the annual report and also computes and plots the isopleths contained in the annual report (see Section 7.2).
8.3 ENVIRONMENTAL CONTRACTOR i
The environmental contractor collects environmental samples and performs radiological analyses as specified in the station's O
8-2 l
1
REVISION O.K JANUARY 1993 radiological environmental monitoring program (see Chapter
()
11).
The contractor issues monthly reports of results to NSEP.
The contractor participates in an Interlaboratory Comparison Program and reports results in the Annual Radiological Environmental Operating Report.
The contractor performs the annual land use census.
The contractor also assembles the Annual Radiological Environmental Operating Report and submits it to NSEP.
8.4 CORPORATE DEPARTMENTS The NSEP administers the offsite dose assessment program.
The department issues and maintains the ODCM (sometimes with assistance from an engineering consultant) as well as some of the procedures associated with implementation of the ODCM.
The department supervises the meteorological and environmental contractors by receiving and reviewing their periodic reports.
It reviews the monthly radiation dose calculations performed by
()
the stations.
It reviews the Annual Radiological Environmental Operating Report, which is assembled by the environmental contractor, and distributes it on behalf of the station.
The Production Engineering and Information Systems Department writes and maintains the computer program used by the stations for offsite dose calculation and projection.
Instructions for using this program are presented in Reference 1.
e O
8-3
REVISION 0.K JANUARY 1993 40.
Commonwealth Edison Company, Information Relevant to
{'
Keepino Levels of Radioactivity in Effluents to Unrestricted Areas As Low As Reasonably Achievable, ha Salle County Station, Units 1 and 2, June 4, 1976.
41.
U.S. Nuclear Regulatory Commission, Branch Technical Position, Radiological Assessment Branch, Revision 1, November 1979.
(This is a branch position on Regulatory Guide 4.8.)
42.
Commonwealth Edison Company, Technical Services Emeroency Plannino Department, Distribution of Monthly, Semiannual t
and Annual Radioloaical and Meteorolocical Reports, EP-ADMIN-7.
43.
U.S. Nuclear Regulatory Commission, Calculation of Releases of Radioactive Materials in Gaseous and Licuid Effluents from Pressurized Water Reactors (PNR-GALE Code),
NUREG-0017, April 1976.
44.
U.S. Nuclear Regulatory Commission, Calculation of i
Releases of Radioactive Materials in Gaseous and Licuid Effluents from Boilina Water Reactors (BWR-GALE Code),
NUREG-0016, April 1976.
45.
Sargent & Lundy, N-16 Skyshine from BWR Turbine Systems and Pipina, NSLD Calculation No. D2-2-85, Rev.
O, 2/1/85.
t 45a. Sargent & Lundy Calculation ATD-0138, Rev.
O, "N-16 Skyshine Ground Level Dose from Dresden Turbine Systems and Piping," July 14, 1992.
45b. Sargent & Lundy Calculation ATD-0139, Rev.
O, "N-16 Skyshine Ground Level Dose from LaSalle Turbine Systems and Piping," July 28, 1992.
45c. Sargent & Lundy Calculation ATD-0140, Rev.
O, "N-16 Skyshine Ground Level Dose from Quad Cities Turbine Systems and Piping," July 28, 1992.
46.
U.S. Nuclear Regulatory Commission, Methods for l
Demonstratina LWR Compliance with the EPA Uranium Fuel 1
l Cycle Standard (40 CFR Part 190), NUREG-0543, i
February 1980.
47.
International Commission on Radiological Protection, Report of Committee Two on Permissible Dose for Internal Radiation, Recommendations of the International Commission on Radiological Protection, ICRP Publication 2, 1959.
48.
U.S. Nuclear Regulatory Commission, Ace-Specific Radiation Dose Commitment Factors for a One-Year Chronic Intake,
()
Battelle Pacific Northwest Laboratories, NUREG-0172, 1977.
R-5
REVISION 0.K JANUARY 1993 49.
W.
C. Ng, Transfer Coefficients for Prediction of the Dose O'
to Man via the Forace-Cow-Milk Pathway from Radionuclides Released to the Biosphere, UCRL-51939.
1 l
50.
E. C.
Eimitis and M. G. Konicek, Derivations of Continuous l
Functions for the Lateral and Vertical Atmospheric Dispersion Coefficients, Atmospheric Environment 6, 859 (1972).
51.
D.
C. Kocher, Editor, Nuclear Decav Data for Radionuclides Qccurrino in Routine Releases from Nuclear Fuel Cycle Facilities, ORNL/NUREG/TM-102, August 1977.
52.
R.
L.
Heath, Gamma-Ray Spectrum Cataloa, Aerojet Nuclear Co., ANCR-1000-2, third or subsequent edition.
53.
S. E. Thompson, Concentration Factors of Chemical Elements in Edible Acuatic Oroanisms, UCRL-50564, Rev.
1, 1972.
54.
U.S. Nuclear Regulatory Commission, Instruction Concernino Risks from Occupational Radiation Exposure, Regulatory Guide 8.29, July 1981.
55.
Dresden Nuclear Power Station, Radioactive Waste and Environmental Monitorino, Annual Report 1987, March 1988.
l 56a. Sargent & Lundy Calculation ATD-0173, Rev.
O, Dated 9/21/92, " Annual Dose to Members of the Public Due to the LaSalle IRSF."
l 56b. Sargent & Lundy Calculation ATD-0174, Rev.
O, Dated 9/21/92, " Annual Dose to Members of the Public Due to the Zion IRSF."
l
~
56c. Sargent & Lundy Calculation ATD-0175, Rev.
O, Dated 9/21/92, " Annual Dose to Members of the Public Due to the Quad Cities IRSF."
56d. Sargent & Lundy Calculation ATD-0176, Rev.
O, Dated 9/21/92, " Annual Dose to Members of the Public Due to the Dresden IRSF."
57a. Sargent & Lundy Calculation ATD-0180, Rev.
O, Dated l
9/25/92, " Dose Information Around Braidwood DAW Sea / Land l
Van Storage Area."
57b. Sargent & Lundy Calculation ATD-0181, Rev.
O, Dated 9/25/92, " Dose Information Around Byron DAW Sea / Land Van Storage Area."
l l
57c. Sargent & Lundy Calculation ATD-0182, Rev.
O, Dated 9/25/92, " Dose Information Around Dresden DAW Sea / Land Van O,
Storage Area."
57d. Sargent & Lundy Calculation ATD-0183, Rev.
O, Dated 9/25/92, " Dose Information Around LaSalle DAW Sea / Land Van Storage Area."
R-6
P REVISION 0.K l
JANUARY 1993 59.
Reserved reference number.
60.
D. C. Kocher, Radioactivity Decay Data Tables, DOE / TIC-11026, 1981.
61.
J. C. Courtney, A Handbook of Radiation Shieldino Data, ANS/SD-76/14, July 1976.
62.
Commonwealth Edison Company, Information Relevant to Keepino Levels of Radioactivity in Effluents to Unrestricted Areas As Low As Reasonably Achievable, Zion Station, Units 1 and 2, June 4, 1976.
63.
Commonwealth Edison Company, Information Relevant to Keepino Levels of Radioactivity in Effluents-to i
Unrestricted Areas As Low As Reasonably Achievable, l
Dresden Station, Units 2 and 3, June 4, 1976.
64.
Commonwealth Edison Company, Information Relevant to Keepino Levels of Radioactivity in Effluents to l
l Unrestricted Areas As Low As Reasonably Achievable, Ouad-Cities Station, Units 2 and 3, June 4, 1976.
l 65.
Sargent & Lundy, METWRSUM, S&L Program Number 09.5.187-1.0.
66.
Sargent & Lundy, Comments on CECO ODCM and List of S&L l
Calculations, Internal Office Memorandum, P.
N.
Derezotes to G. R.
Davidson, November 23, 1988.
67.
Sargent & Lundy, AZAP, A Computer Procram to Calculatg Annual Averace Offsite Doses from Routine Releases of Radionuclides in Gaseous Effluents and Postaccident X/O Values, S&L Program Number 09.8.054-1.7.
68.
National Oceanic and Atmospheric Administration, A Procram
~
for Evaluatino Atmospheric Dispersion from a Nuclear Power Station, J. F. Sagendorf, NOAA Technical Memorandum ERL ARL-42, Air Resources Laboratory, Idaho Falls, Idaho, May 1974.
O R-7 I
i I'
REVISION 0.K JANUARY 1993 O
69.
G.
P. Lahti, R.
S. Hubner, and J.
C. Golden, Assessment of Gamma-Rav Exposures Due to Finite Plumes, Health Physics 41, 319 (1981).
i 70.
Reserved reference number.
71.
Reserved reference number.
72.
W.
R. Van Pelt (Environmental Analysts, Inc.), Letter to J. Golden (CECO) dated January 3, 1972.
l 73.
Electric Power Research Institute, Radiolooical Effects of Hydrocen Water Chemistry, EPRI NP-40ll, May 1985.
74.
U.S. Nuclear Regulatory Commission, Draft Generic Environmental Impact Statement on Uranium Millina, NUREG-0511, April 1979.
l 75.
U.S. Environmental Protection Agency, Environmental Analysis of the Uranium Fuel Cycle, Part I - Fuel Supply, EPA-520/9-73-003-B, October 1973.
76.
U.S. Nuclear Regulatory Commission, Final Generic Environmental Statement on the Use of Recycle Plutonium in Mixed Oxide Fuel in Licht Water Cooled Reactors, NUREG-0002, August 1976.
77.
U.S.
Nuclear Regulatory Commission, Democraphic Statistics Pertainina to Nuclear Power Reactor Sites, NUREG-0348, Draft, December 1977.
I 78.
Nuclear News 31, Number 10, Page 69 (August 1988).
79.
General Electric Company, Irradiated Fuel Storace at Morris Operation. Operatino Experience Report, January 1972 throuch December 1982, K. J.
Eger, NEDO-20969B.
80.
U.S. Nuclear Regulatory Commission, Generic Letter 89-01,
" Guidance For The Imulementation of Procrammatic Controls For RETS In The Administrative Controls Section of Technical Specifications and the Relocation of Procedural Details of Current RETS to the Offsite Dose Calculation Manual or Process Control Procram", January 1989,.
81.
" Assessment of the Impact of Licuid Radioactive Effluents from Braidwood Station on Proposed Public Water Intakes at Wilmincton, Illinois", J.C. Golden, NSEP, January 1990.
82.
NRC Safety Evaluation Report (SER)/ Idaho National Engineering Laboratory Technical Evaluation Report (TER) of the Commonwealth Edison Offsite Dose Calculation Manual
()
(ODCM), Revision O.A, December 2, 1991.
R-8
REVISION 0.K JANUARY 1993
)
is classified as one of the following three height-dependent types, which are defined in Section 4.1.4:
Stack (or Elevated) Release Point (denoted by the symbol S or subscript s)
Ground Level Release Point (denoted by the symbol G or subscript g)
Vent (or Mixed Mode) Release Point (denoted by the symbol V or subscript v)
The release point classifications of routine release points at the nuclear stations are stated in Table A-1 of this appendix.
i Releases from points not listed in Table A-1 are considered ground level releases.
A.l.2 Dose Due to Noble Gas Radionuclides A.l.2.1 Gamma Air Dose I
1
(
Requirement l
l l
Standard Tec:r? cal Specifications (References 2 and 3) limit the gamma air dose due to noble gas effluents released from each reactor unit to area at and beyond the site boundary to the following (see Specification 3.11.2.2):
Less than or equal to 5 mrad per calendar quarter.
Less than or equal to 10 mrad per caIOndar year.
This provision is related to one of the design objectives of l
l l
I Section 12.4 of each station's RETS contain a somewhat similar l
provision.
- CE) l A-2 l
i 4
REVISION 0.K JANUARY 1993
(
Equation 1
The gamma air dose due to noble gases released in gaseous effluents is calculated by the following expression:
Dy = (3.17E-8)E{ S Ai is + V Aiy + GAig }
(A-1) i i
The summation is over noole gas radionuclides i.
Dy Gamma Air Dose
[ mrad]
Dose to air due to gamma radiation from noble gas radionuclides released l
in gaseous effluents.
3.17E-8 Conve,sion Constant
[yr/sec) i l
Conv' arts seconds to years.
l i
Si, Vi, Gi Gamna Air Dose Factor
[(mrad /yr)/
j (pCi/sec)]
I Gamma air dose rate at a specified location l
per unit of radioactivity release rate for radionuclide i released from a stack, vent, l
or ground level release point, respec-l(
tively.
See Section 4.2.1.1, Section B.5 of Appendix B, and Table F-7 of Appendiz F.
l Ais, Aiv, Aig Cumulative Radionuclide Release
[pCi]
l l
Measured cumulative release of radionuclide i over the time period of interest from a l
stack, vent, or ground level release point.
m O
A-3
i l
REVISION 0.K j
JANUARY 1993 O
Application Standard Technical Specifications (References 2 and 3) require determination of cumulative gamma air dose contributions due to noble gases for the current calendar quarter and the current calendar year at least once per 31 days (see Specification 4.11.2.2).
Sections 12.4 of each station's RETS contain a somewhat similar provision.
I The dose factors in Table F-7 of Appendix F are used for the j
determinations required by these specifications.
These values l
were calculated for the site boundary.in each sector and are
{
judged to be very good approximations to the maximum offsite l
values.
After doses for all sectors are determined, the highest dose is compared with the RETS limit on gamma air dose.
I For a release attributable to e processing or effluent system shared by more than one reactor unit, the dose due to an indivi-i dual unit is obtained by proportioning the effluents among the j
units sharing the system.
The allocation procedure is speci-l fied in Chapter 10 of this manual.
l i
~
1
! O A-4 l
REVISION 0.K JANUARY 1993
()
A.l.2.2 Beta Air Dose Requirement Standard Technical Specifications (References 2 and 3) limit the beta air dose due to noble gases in gaseous effluents released from each reactor unit to areas at and beyond the site boundary to the following (see Specification 3.11.2.2):
Less than or equal to 10 mrad per calendar quarter.
Less than or equal to 20 mrad per calendar year.
This provision is related to one of the design objectives of 10 CFR 50, Appendix I.
Section 12.4 of each station's RETS contain a somewhat similar provision.
O Equation The beta air dose due to noble gases released in gaseous effluents is calculated by the following expression:
a
= (3.17E-8)L{
Li[(X/Q)sA'is + (X/Q)yA'iy (A-2)
D (X/Q)gA'ig] }
+
O A-5
(.
REVISION 0.K JANUARY 1993 exp(-l R/3600u )
(A-5)
A'ig = Aig i
g i
- Ais, Cumulative Radionuclide
[pCi]
l
- Aiy, Release Aig Defined in Section A.l.2.1.
Ai Radiological Decay Constant
[hr-1]
Radiological decay constant for radionuclide i.
See l
Table D-12 of Appendix D.
l l
R Downwind Range
[m]
Distance from the release point to the dose point.
See Tables F-5, F-6, and F-7.
3600 Conversion Constant
[sec/hr]
Converts hours to seconds.
u, Average Wind Speed
[m/sec]
s uy,ug Average wind speed for a stack, vent, or ground level i
x release.
See Section B.l.3 of Appendix B and Table F-4 of Appendix F.
Application Standard Technical Specifications (References 2 and 3) require determination of cumulative beta air dose contributions due to noble gases for the current calendar quarter and the current calendar year at least once per 31 days (see Specification 4.11.2.2).
Section 12.4 of each station's RETS contain a
~
somewhat similar provision.
A-7
REVISION 0.K JANUARY 1993 Beta air dose is determined for each sector using the highest
()
calculated offsite value of X/O for that sector.
This value and the distance R to which it pertains are provided in Table F-5 of Appendix F.
The highest dose is compared with the Tech ncal Specifications limit on beta air dose.
l For a release attributable to a processing or effluent system shared by more than one reactor unit, the dose due to an indivi-j dual unit is obtained by proportioning the effluents among the units sharing the system.
The allocation procedure is speci-fied in Chapter 10 of this manual.
A.l.2.3 Whole Body Dose Requirement 1
Standard Technical Specifications (References 2 and 3) limit the annual dose commitment to any member of the public due to
()
releases of radioactivity and to radiation from uranium fuel cycle sources to less than or equal to 25 mrem to the whole body (see Specification 3.11.4).
The limit applies to the sum of doses due to radioactive effluents (airborne and liquid) and doses due to direct radiation from noneffluent sources (e.g.,
sources contained in systems on site).
This specification is related to a dose limit of 40 CFR 190.
See Section A.3 for further discussion.
Section 12.4 of each station's RETS contain a somewhat similar provision.
O A-8
REVISION 0.K JANUARY 1993
("}
Equation U
The part of whole body dose due to gamma radiation from noble gases released in gaseous effluents is calculated by the follow-ing expression:
Dwd = (0.7)(1.ll)(3.17E-8)
(A-6) x E{ 5 A1 is +
iAiy + CAi ig }
The summation is over noble gas radionuclides i.
Dwb Whole Body Dose
[ mrem]
Dose to the whole body due to gamma radiation from noble gas radionuclides released in gaseous effluents.
0.7 Shielding Factor Dimensionless factor which accounts for shielding due to occupancy of structures.
1.11 Conversion Constant
[ mrem / mrad]
Converts rads in air to rems in tissue.
3.17E-8 Conversion Constant
[yr/sec]
Converts seconds to years.
5,V,Ci Gamma Whole Body Dose Factor
[(mrad /yr)/
1 i
(pCi/sec)]
Gamma whole body dose rate at a specified location per unit of radioactivity release rate for radionuclide i released from a stack, vent, or ground level release point.
O A-9
REVISION 0.K JANUARY 1993 The attenuation of gamma radiation due to passgge through 5 cm of body tissue of 1 l
g/cm density is taken into account in calculating this quantity.
See Section 4.2.1.3, Section B.6 of Appendix B, and Table F-7 of Appendix F.
{
Cumulative Radionuclide Release
[pCi]
Ais, Aiy, Aig Defined in Section A.l.2.1.
]
l Application The standard Technical Specifications require calculation of l
whole body dose due to airborne effluents only in connection with assessment of compliance with 40 CFR 190.
Because of the low dose limits associated with 10 CFR 50, stations are not required to assess compliance with 40 CFR 190 unless a 10 CFR 50 limit is exceeded by a specified amount (see Section A.3).
When 40 CFR 190 assessments are made, whole body doses due to other sources (e.g.,
inhalation, ingestion, ground radiation 1
and direct radiation from contained sources) must also be considered.
See Section A.3.1 for further discussion.
(
A.l.2.4 Skin Dose Requirement There is no regulatory requirement to evaluate skin dose.
However, this component is evaluated for reference as there is a skin dose guideline contained in 10CFR50, Appendix I.
In the l
unlikely event where the beta air dose guideline.is exceeded, the skin dose will require evaluation.
i A-10
REVISION 0.K JANUARY 1993 Application The skin dose is calculated for reference only.
A.l.3 Dose Rate Due to Noble Gas Radionuclides A.1.3.1 Whole Body Dose Rate Requirement i
Standard Technical Specifications (References 2 and 3) limit j
the whole body dose rate due to noble gases in gaseous effluents released from a site to areas at and beyond the site boundary to less than or equal to 500 mrem /yr at all times (see Specification 3.11.2.1).
This provision is related to the requirements of 10 CFR 20.105 and 20.106.
Section 12.4 of each station's RETS contain a somewhat similar provision.
b
=
A-12
REVISION 0.K JANUARY 1993 i
()
Equation i
I The whole body dose rate due to noble gases released in gaseous effluents is calculated by the following expression:
wb = (1 ll)EI 5 Q s + I O v + E 0i ig }
(A-8)
J b
ii ii The summation is over noble gas radionuclides i.
b Whole Body Dose Rate
[ mrem /yr]
wb I
Dose rate to the whole body due to gamma radiation from noble gas radionuclides j
released in gaseous effluents.
i i
Release Rate
[pCi/sec]
Qis, C v, Qig Measured release rate of radionuclide i from a stack, vent, or ground level release point.
The remaining parameters have the same definitions as used in the equation for whole body dose in Section A.1.2.3.
Application Standard Technical Specifications (References 2 and 3) require the dose rate due to noble gases in gaseous effluents to be determined to be within the above limit in accordance with methodology specified in the ODCM (see Specifica' tion 4.11.2.1.1).
The Section 12.4 of each station's RETS contain a i
somewhat similar provision.
l l
l A-13 l
l
i l
REVISION 0.K JANUARY 1993 l
To comply with this specification, each station uses an
)
[}
effluent radiation monitor satpoint corresponding to an offsite whole body dose rate at or below the limit (see Chapter 10).
In addition, each station assesses compliance by calculating offsite whole body dose rate on the basis of samples obtained periodically in accordance with station procedures.
I I
A.l.3.2 Skin Dose Rate j
Requirement Standard Technical Specifications (References 2 and 3) limit the skin dose rate due to noble gases in gaseous effluents j
released from a site to areas at and beyond the site boundary to less than or equal to 3000 mrem /yr at all times (see Specification 3.11.2.1).
This provision is related to requirements of 10 CFR 20.105 and 20.106.
(
Section 12.4 of each station's RETS contain a somewhat similar provision.
l I
s i
O A-14 l
l
REVISION 0.K JANUARY 1993 The remaining parameters have the same definitions as used in l f~
the equation for skin dose in Section A.l.2.4.
(
Application Standard Technical Specifications (References 2 and 3) require l
l the dose rate due to noble gases in gaseous effluents to be I
l determined to be within the above limit in accordance with I
methodology specified in the ODCM (see Specification 4.11.2.1.1).
Section 12.4 of each station's RETS contain a somewhat similar provision.
To comply with this specification, each station uses an effluent radiation monitor setpoint corresponding to an offsite skin dose rate at or below the limit (see Chapter 10).
In addi-tion, each station assesses compliance by calculating offsite skin dose rate on the basis of samples obtained periodically in accordance with station procedures.
O i
l l
O A-16
t REVISION 0.K 4
JANUARY 1993 I
A.l.4 Dose Due to Non-Noble-Gas Radionuclides Requirement Standard Technical Specifications (References 2 and 3) provide the following limits on the dose to a member of the public from specified non-noble-gas radionuclides in gaseous effluents released from each reactor unit to areas at and beyond the site-boundary (see Specification 3.11.2.3):
Less than or equal to 7.5 mrem to any organ during any e
calendar quarter.
Lass than or equal to 15 mrem to any organ during any calendar year.
This provision is related to one of the design objectives of 10 CFR 50, Appendix I.
Section 12.4 of each station's RETS contain a somewhat similar provision.
t i
e O
A-17 i
REVISION 0.K JANUARY 1993
()
In the standard specification (References 2 and 3) and the specifications of all statiores except LaSalle and Zion, the i
i specified non-noble-gas radionuclides are Iodine-131.
Iodine-133.
All radionuclides in particulate form with half-lives greater than 8 days.
For LaSalle and Zion, the specified radionuclides are:
Radiciodines with half-lives greater than 8 days.
Radioactive materials in particulate form with half-lives greater than 8 days.
Radionuclides, other than noble gases,.with half-lives greater than 8 days.
O This section provides expressions used to calculate dose to an organ due to non-noble-gas radionuclides released in gaseous l
effluents.
The following organs are considered:
Total body.
Bone.
Liver.
Thyroid.
Kidney.
Lung.
GI-LLI (gastrointestinal tract and lower large intestine).
i O
A-18
- ~..
REVISION 0.K JANUARY 1993 Equation The dose calculated includes dose commitment (see Section 4.1.1) incurred due to teleases in the time period under con-sideration.
Specifically, dose is calculated as the sum of three contributions:
ja = Dgndj + D nhalja + D ja (A-13) i f
d NNG D
NNG Dose Due to Non-Noble-Gas
[ mrem]
D ja Radionuclides Sum of the dose and dose commitment to organ j of an individual of age group a due j
to non-noble-gas radionuclides released in gaseous effluents during a specified time period.
Dgnd Ground Deposition Dose
[ mrem]
j Dose to organ j due to ground deposition of j
non-noble-gas radionuclides released in gaseous effluents.
See Equation A-14 in Section A.1.4.1.
i D nhal Inhalation Dose
[ mrem]
ja j
Dose commitment to organ j of an individual of age group a due to inhalation of non-l noble-gas radionuclides released in gaseous l
effluents.
See Equation A-17 in Section j
A.l.4.2.
f d
~
D ja Food pathways Dose
[ mrem]
Dose commitment due to ingestion via food pathways (leafy vegetables, produce, milk, and meat) of non-noble-gas radionuclides released in gaseous effluents.
See Equation l
A-18 in Section A.l.4.3.
Application Standard Technical Specifications (References 2 and 3) require cumulative ~ dose contributions for the current calendar quarter and the current calendar year for the specified non-noble-gas radionuclides in airborne effluents to be determined at least i
A-19
REVISION O.K JANUARY 1993 once per 31 days (see Specification 4.11.2.3).
Section 12.4 of
(
each station's RETS contain a somewhat similar provision.
To comply with this specification, each station obtains and analyzes samples in accordance with the radioactive gaseous waste or gaseous effluent sampling and analysis program in its Technical Specifications.
For each organ of each age group considered, the dose for each pathway is calculated in every sector (except Zion sectors over Lake Michigan).
The calculation is based on the location assumptions discussed below in conjunction with the pathway equations.
For each organ of each age group, the doses are summed in each sector over all pathways (ground, inhalation, and four food pathways).
The result for the sector with the highest total dose is compared to the limit.
For a release attributable to a processing or effluent system shared by more than one reactor, the dose due to an individual
()
unit is obtained by proportioning the effluents among the units sharing the system.
The allocation procedure is specified in Chapter 10 of this manual.
b o
O I
A-20 i
i
REVISION 0.K JANUARY 1993 A.l.4.2 Inhalation The dose commitment due to inhalation is claculated by the following expression:
iD nhalja = (3.17E-8)(lE6)(R )
(A-17) a x E{ DFA jaE(X/Q)sA*is + (X/Q)yA'iy i
(X/Q)gA'ig] }
+
The summation is over non-noble-gas radionuclides i.
i D nhal Inhalation Dose Commitment
[ mrem]
ja Dose commitment to organ j of an individual in age group a due to inhalation of non-noble-gas radionuclides released in gaseous effluents.
3.17E-8 Conversion Constant
[ yrs /sec]
Converts seconds to years.
()
lE6 Conversion Constant
[pCi/pCi]
Converts pCi to pCi.
3 R
Individual Air Intake Rate
[m /yr]
a Air intake rate for individuals in age group a.
See Table D-7 Appendix D.
DFA ja Inhalation Dose Commitment Factor
[ mrem /pCi]
i Dose commitment to organ j of an individual in age group a per unit of activity of radionuclide i inhaled.
See Tables D-1 through D-3 Appendix D.
3 (X/Q)s, Relative Effluent Concentration
[sec/m )
(X/Q)y, (X/Q)g Radioactivity concentration at a specified location per unit of radioactivity release rate.
See Section 4.1.6, Section B.3 of Appendix B, and Table F-1 of Appendix F.
O A-23
REVISION 0.K JANUARY 1993 DFA ja Ingestion Dose Commitment Factor
[ mrem /pCi]
i Dose commitment to organ j of an individual t
l in age group a per unit of activity of radio-I nuclide i ingested.
See Tables D-4, D-4a, D-4b and D-5 of Appendix D.
iV
,iP Rate of Ingestion of Activity
[pCi/yr]
ia, iMia'iFia Activity of radionuclide i ingested annually by an individual in age group a from, respec-tively, the following:
Leafy vegetables.
Produce (nonleafy vegetables, l
fruits, and grain).
l Milk.
1 Meat (flesh).
l l
Calculated as follows:
V V
iV
=U I
C i (g_yg) a V
iP
=U fp C i (A-20)
P P
a iMia = U a M
M Ci (A-21) l F
F F
Ci (A-22)
I
=U a
V U
Food Product Consump-
[kg/yr]
a tion Rate UP
[kg/yr]
a M
U IL/YTl a
F U
[kg/yr]
a Annual consumption (usage) rate of leafy vegetables, produce, milk, or meat, respectively, for individuals in age group a.
See Table D-7 of Appendix D.
1 j
fy Food Product Affected Fraction fp Fraction of ingested leafy vegetables (V) or produce (P) grown in the garden of interest.
See Table D-6 of i
l Appendix D.
A-25 l
REVISION 0.K JANUARY 1993 V
C i Food Product Radioactivity
[pCi/kg]
)
Concentration 1
P C i
[pCi/kg]
M C i
[pCi/L]
F C i
[pCi/kg]
V P
C i and C i represent, respectively, the average concentration of radionuclide i in leafy vegetables and produce grown in the garden of interest.
Calculated from the amount of radioactivity released and the relative deposition factor D/Q at the garden of interest.
See Section A.1.4.3.1 below I
M F
for the equation.
C i and C i represent, respectively, the average concentration of radionuclide i in milk and meat from the
)
producer of interest.
Calculated from the amount of radioactivity released and the relative deposition factor D/Q at the locations of the producers of interest.
See Sections A.l.4.3.2 and A.l.4.3.3 below for I
equations.
Application The dose due to ingestion of leafy vegetables and produce is calculated in each sector for a hypothetical garden assumed to (d
be located at the location of highest offsite D/O (see Table s
F-5 of Appendix F).
The dose due to ingestion of milk and meat is calculated in each sector for the location of the nearest producer as specified in Table F-6 of Appendix F.
If there is no actual milk or meat producer within 5 miles of the station, one is assumed to be located at 5 miles (except that no food pathway calculations are made for Zion sectors in which the offsite regions near the station are over Lake Michigan).
A.1.4.3.1 Vegetation The radioactivity concentration in leafy vegetables ('CV ),
pro-i P ), or other vegetation is calculated by the following duce (C i
expression:
L)s A-26
REVISION 0.K JANUARY 1993
)
[(d )(r)/(Y )(AEi)]
(A-23)
Ci=
i y
()
[exp(l t )](fg) x [1 - exp(-1Ei e)]
t ih l
Ci Food Product Radioactivity
[pCi/kg]
Concentration Average concentration of radionuclide i in leafy vegetables, produce, or other vegetation.
2 di Deposition Rate
[(pCi/hr)/m )
Rate at which radionuclide i is deposited onto the ground.
Calculated from the amount of radioactivity released and the relative deposition factor D/Q at the location of interest.
See Section A.l.4.1 for an equa-tion.
See the Subsection " Application" in Section A.1.4.3 for the location assumptions j
used in determining d,
i r
Vegetation Retention Factor Fraction of deposited activity retained on vegetation.
See Table D-6 of Appendix D.
2 Y
Agricultural Productivity
[kg/m )
y Yield The quantity of vegetation produced per unit area of the land on which the vegetation is grown.
AEi Effective Decay Constant
[hr-1]
Effective removal rate constant for radio-nuclide i from vegetation:
AEi
- Ai+A (A-24) w li Radiological Decay Constant
[hr-1]
Radiological decay cons' tant for radionuclide i.
See Table D-12 of Appendix D.
l Weathering Decay Constant
[hr-1]
y Removal constant for physical loss by weathering.
See Table D-6 of Appendix D.
{
($)
A-27
REVISION 0.K JANUARY 1993 t
Effective Vegetation
[hr]
e Exposure Time Time that vegetation is exposed to contamination during the growing season.
See Table D-6 of Appendix D.
Harvest to Consumption
[hr]
th Time Time between harvest and consumption.
See Table D-6 of Appendix D.
fg Seasonal Growing Factor Factor which accounts for the seasonal j
growth of vegetation.
It has the value 1 during the growing season, O otherwise.
See Table D-6 of Appendix D.
A.1.4.3.2 Milk i
The radioactivity concentration in milk is calculated by the following expressions:
M f
l C i=F C i Wf exp(-l ti M)
(A-25) f a
g i+ (1 - f )Csi + fa(1 - fg)Csi (A-26)
C i =f f
C9 a
M C i Milk Radioactivity Concentration
[pCi/L]
Average concentration of radionuclide i in milk from the producer of interest.
F Milk Fraction
[ days /L]
M Fraction of an animal's daily intake of radionuclide i which appears in each liter of milk (pCi/L in milk._per pCi/ day ingested by the animal).
See Table D-8 of Appendix D.
f Feed Concentration
[pCi/kg]
C i Average concentration of radionuclide i in animal feed.
A-28
REVISION 0.K JANUARY 1993 f
Feed Consumption
[kg/ day)
W Amount of feed consumed by the animal O
each day.
See Table D-6 of Appendix D.
Ai Radiological Decay Constant
[hr-1]
Radiological decay constant for radio-nuclide i.
See Table D-12 of Appendix D.
t Milk Transport Time
[hr]
M Average time from the production of milk to its consumption.
See Table D-6 of Appendix D.
f pasture Time Fraction a
Fraction of time that animals graze on pasture.
See Table D-6 of Appendix D.
f Pasture Grass Fraction g
Fraction of daily feed that is pasture grass when animals graze on-pasture.
See Table D-6 of Appendix D.
C9 Pasture Grass Concentration
[pCi/kg]
1 Concentration of radionuclide i in pasture grass.
Calculated using Equation A-23 with the seasonal growing factor ff = 1 and with O
parameter values specified for the pasture grass and milk pathways in Table D-6 of Appendix D.
CS Stored Feed Concentration
[pCi/kg]
i Concentration of radionuclide i in stored feed.
Calculated using Equation A-23 for Ci with the seasonal growing factor ff=1 and parameter values specified for the stored feed and milk pathways in Table D-6 of Appendix D.
A.1.4.3.3 Meat s
The radioactivity concentration in meat is calculated by the following expression:
F f
C i = Fy C i Wf exp(-l t )
(A-27) is A-29
~.
REVISION 0.K JANUARY 1993 F
C i Meat Radioactivity Concentration
[pCi/kg]
[~
Average concentration of radionuclide
(-
i in meat from the producer of interest.
Fp Meat Fraction
[ days /kg]
Fraction of an animal's daily intake of radionuclide i which appears in each kilo-gram of flesh (pCi/kg in meat per pCi/ day ingested by the animal).
See Table D-8 of Appendix D.
f Feed Concentration
[pCi/kg]
C i Average concentration of radionuclide iinanimalfegd.
Calculated using the equation for C i in the preceding sub-section with parameter values specified for the meat pathway in Table D-6 of Appendix D.
g Feed Consumption
[kg/ day]
W Amount of feed consumed by the animal each day.
See Table D-6 of Appendix D.
i Radiological Decay Constant
[hr-lj A
Radiological decay constant for radio-i nuclide i.
See Table D-12 of Appendix D.
/~
t Slaughter to Consumption Time
[hr]
s Time from slaughter to consumption.
See Table D-6 of Appendix D.
A.1.5 Dose Rate Due to Non-Noble-Gas Radionuclides Requirement Standard Technical Specifications (References 2 and 3) limit the dose rate to any organ due to radioactive materials in i
gaseous effluents released from a site to areas at and beyond the site boundary to less than or equal to 1500 mrem /yr (see Specification 3.11.2.1).
This provision is related to the requirements of 10 CFR 20.105 and 20.106.
The Bases section of the Standard Technical Specifications states that it restricts the thyroid dose rate to a child via the inhalation pathway to 1500 mrem /yr.
A-30 i
REVISION 0.K JANUARY 1993 Section 12.4 of each station's RETS contain somewhat similar
()
provisions.
In accordance with the guidance in the Bases section of their RETS, all stations consider the child to be the receptor in calculating dose commitment to the thyroid due to inhalation of
)
non-noble-gas radionuclides in gaseous effluents.
Each station calculates doses for all members of the public then determines the maximum dose.
The member of the public who receives the maximum dose will be reported.
In the standard specifications and the RETS of all stations except LaSalle and Zion, the specified non-noble-gas radionuclides are Iodine-133.
Iodine-133.
All radionuclides in particulate form with half-lives e
greater than 8 days.
For Zion, the specified radionuclides are as above except that tritium is omitted.
A-31
i REVISION 0.K JANUARY 1993 For LaSalle, the specified radionuclides are Radioiodines with half-lives greater than 8 days.
Radioactive materials in particulate form with half-lives greater than 8 days.
Radionuclides, other than noble gases, with half-lives greater than 8 days.
i l
Equation 1
The dose rate to a child thyroid due to inhalation is calcu-lated by the following expression:
i b nhalja = (lE6)(R )E{ DEA ja[(X/Q)30*is (A-28) a i
(X/Q)gQ*ig] }
(X/Q)yQ'iy
+
+
The summation is over non-noble-gas radionuclides i.
O i
b nhal Inhalation Dose Rate
[ mrem /yr]
ja Rate of dose commitment to organ j of an individual in age group a due to inhalation of non-noble-gas radionuclides released in gaseous effluents; j and a are chosen to
~
correspond to a child thyroid.
Q'is.
Radionuclide Release Rate,
[pCi/sec]
Q*iv.
Adjusted for Radiodecay O'ig Measured release rate of radionuclide i from a stack, vent, or ground level release point, reduced to account for radio-decay in transit from the release point to the dose point.
See Section A.l.3.2.
The other parameters are defined in Section A.l.4.2.
rm N
A-32
REVISION 0.K JANUARY 1993 Application O
Standard Technical Specifications (References 2 and 3) require the dose rate due to non-noble-gas radioactive materials in air-borne effluents to be determined to be within the above limit in accordance with a sampling and analysis program specified in the Technical Specifications (see Specification 4.11.2.1.2).
Section 12.4 of each station's RETS contain a somewhat similar provision.
To comply with this specification, each station obtains and analyzes samples in accordance with the sampling and analysis program in its RETS.
The child thyroid dose rate due to l
inhalation is calculated in each sector at the location of the highest offsite X/Q.
The result for the sector with the l
highest child thyroid inhalation dose rate is compared to the limit.
()
A.1.6 Operability and Use of Gaseous Effluent Treatment Systems l
Requirement Standard Technical Specifications for a pressurized water j
reactor (References 2 and 3) require that the ventilation exhaust l
t l
l O
A-33 l
t
REVISION 0.K JANUARY 1993 I
treatment system and the waste gas holdup system be used when projected offsite doses in 31 days due to gaseous effluent
[
releases, from each reactor unit, exceed any of the following l
limits (see Specification 3.11.2.4):
l r
I e
0.2 mrad to air from gamma radiation, j
i e
0.4 mrad to air from beta radiation.
0.3 mrad to any organ of a member of the public.
e This provision is related to the requirements of 10 CFR 50, Appendix I.
t The RETS of some stations contain a somewhat similar provision.
For exact requirements, see the following sections:
l e
Braidwood 1/2:
12.4.4 e
Byron 1/2:
12.4.4 e
La Salle 1/2:
12.4.5 These stations are required to project doses due to gaseous releases from the site at least once per 31 days.
In addition, Dresden 2.3 and Quad Cities 1/2 are required to operate the off-gas treatment system at certain times.
In conjunction with this requirement, Dresden and Quad Cities are required to project offsite doses due to gases treated by the off-gas treatment system at least once per 31 days (see Dresden and Quad Cities Section 12.4.4).
O A-34
REVISION 0.K JANUARY 1993 Equation O
Offsite doses due to projected releases of radioactive j
materials in gaseous effluents are calculated using Equations A-1, A-2, and A-13.
Projected cumulative radionuclide releases are used in place of measured cumulative releases Ais, Aiy, and Aig.
Application l
l For a release attributable to a processing or effluent system l
shared by more than one reactor unit, the dose due to an indivi-dual unit is obtained by proportioning the effluents among the l
units sharing the system.
The allocation procedure is speci-fied in Chapter 10 of this manual.
A.2 LIQUID RELEASES (10 CFR 20 AND 10 CFR 50, APPENDIX I)
(~
\\
A.2.1 Dose i
i Requirement i
Standard Technical Specifications (References 2 and 3) provide the following limits on the dose or dose commitment to a member of the public from radioactive materials in liquid effluents released from each reactor unit to unrestricted areas (see Specification 3.11.1.2):
s During any calendar quarter, less than or equal to 1.5 e
mrem to the whole body and less than or equal,to 5 mrem
)
l to any organ.
During any calendar year, less than or equal to 3 mrem to the whole body and less than or equal to 10 mrem to any organ.
This provision is related to one of the design objectives of 10 l
A-35 l
l
~
REVISION 0.K JANUARY 1993 Section 12.3 of each station's RETS contain a somewhat similar l
provision.
Equation The dose commitment from radioactive materials in liquid efflu-ents is calculated for all age groups.
The dose commitment is obtained as the sum of contributions from consumption of drinking water and fish:
fish D1I9
= Dwaterja + D ja (A-29) ja Dwaterja= (1.1E-3)(8760)(U M /F )
(A-30)
W W W
x E{ A DFIijaexp(-l t"))
j i
i fish Ef f
D ja = (4.lE-3)(8760)(U M /F )
(A-31) f x E{-A B DFI jaerp(-l t ))
ii i
i The summations are over index i (radionuclides).
Dli9 Dose Commitment Due to Radioactivity
[ mrem]
ja in Liquid Effluents Dose commitment to organ j of age group a consuming water and fish containing radio-activity released in liquid effluents.
O O
I A-36 l
-. ~.., - - - -. -,
REVISION 0.K JANUARY 1993 Dwater Dose Commitment Due to Consumption
[ mrem]
ja l
of Drinking Water j (
Dose connitment to organ j of age group a consuming water containing radio-I activity released in liquid effluents.
fish Dose Commitment Due to Consumption
[rcrem]
D ja of Fish Dose commitment to organ j of age group a consuming fish containing radioactiv.ty released in liquid effluents.
f Usage Factor
[L/hr, kg/hr]
i UW' U a
a W
f l
Consumption rate of water (U ) or fish (U ).
a See Table D-7 of Appendiz D.a l
W f
1/M, 1/M Dilution Factor Measure of dilution prior to withdrawal of potable water or fish.
See Table F-1 of Appendix F.
W F
Average Flow Rate
[cfs]
Average flow rate of receiving body of water I
at point where potable water is taken.
See l
Table F-1 of Appendix F.
I f
F Near-Field Flow Rates
[cfs]
Near-field flow rate of receiving body of water (in region where fish are taken).
See l
Table F-1 of Appendix F.
Ai Radionuclide Release (pCi]
Measured amount of radionuclide i released
~
in liquid effluents during the time period under construction.
j DFI j a Ingestion Dose Factor
[ mrem /pCi]
i Dose commitment to organ j of an individual in age group a per unit of activity of radionuclide i ingested.
See Table D-4, l
D-4a, D-4b, D-5 of Appendix D.
Ai Decay Constant
[hr-1]
Radiological decay constant of radionuclide i.
See Table D-12 of Appendix D.
I l
O A-37 l
REVISION 0.K JANUARY 1993 W
f t, t Elapsed Time
[hr]
Average elapsed time between release and O
consumption of potable water or fish.
See Table F-1 of Appendix F.
Bioaccumulation Factor
[L/kg]
Bi Equilibrium ratio of the concentration of-radionuclide i in fish (pCi/kg) to its concentration in water (pCi/L).
See Table D-13 of Appendix D.
1.1E-3 Conversion Constant
[(pCi/ liter) per (pCi/yr)/(cfs)]
Factor to convert to pCi/ liter from j
(pCi/yr)/(cfs).
i 8760 Conversion Constant
[hr/yr]
Number of hours per year.
l Application Standard Technical Specifications (References 2 and 3) require determination of cumulative dose contributions from liquid effluents for the current calendar quarter and the current
()
calendar year at least once per 31 days (see Specification 4.11.1.2).
Section 12.3 of each station's RETS contain a somewhat similar provision.
l l
t For a release attributable to a processing or effluent system 1
shared by more than one reactor unit, the dose due to an individual unit is obtained by proportioning the effluents l
among the units sharing the system.
The allocation procedure l
is specified in Chapter 10 of this manual.
l l
A-38 l
l
REVISION 0.K JANUARY 1993 A.2.2 Maximum Permissible Concentration O
Requirement Standard Technical Specifications (References 2 and 3) provide that the concentration of radioactive material released in liquid effluents to unrestricted areas shall be limited as follows (see Specification 3.11.1.1):
e For radionuclides other than dissolved or entrained noble gases, to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2.
For dissolved or entrained noble gases, to 2E-4 pCi/mL e
total activity.
This provision is related to the requirements of 10 CFR 20.106.
Section 12.3 of each station's RETS contain a somewhat similar provision.
O For the concentration of dissolved or entrained noble gases, the RETS of Braidwood and Byron specify the same limit (2E-4 pCi/mL) as the standard Technical Specifications.
The RETS of the remaining stations specify different values (see Table D-ll of Appendix D).
The limit for the concentration of a radionuclide in liquid released to the unrestricted area is called its maximum permissible concentration (MpC).
s e
O A-39
l REVISION 0.K JANUARY 1993 Application O
Standard Technical Specifications (References 2 and 3) require 1
a specified sampling and analysis program to assure that liquid j
radioactivity concentrations at the point of release are maintained within the required limits (see Specifications 4.11.1.1.1 and 4.11.1.1.2).
Section 12.3 of each station's I
RETS contain a somewhat similar provision.
l l
To comply with this provision, each station obtains and analyzes samples in accordance with the radioactive liquid waste (or effluent) sampling-and analysis program in its RETS.
Radioactivity concentrations in tank effluents are determined q
in accordance with Equation A-33 in the next section.
Comparison with the MPC limit is made using Equation A-32.
A.2.3 Tank Discharges i
I
()
When radioactivity is released to the unrestricted area with i
liquid discharge from a tank (e.g.,
a radwaste discharge tank),
the concentration of a radionuclide in the effluent is calculated as follows:
Ci = (Ct )(pr)f(pd, pr)
(A-33) s e
O A-41 9
REVISION 0.K JANUARY 1993 Ci Concentration in Liquid Effluent
[pCi/mL]
i to the Unrestricted Area
()
Concentration of radionuclide i in liquid released to the unrestricted area.
Ct Concentration in the Discharge Tank
[pCi/mL]
i Measured concentration of radionuclide i in the discharge tank.
T F
Flow Rate, Tank Discharge
[cfs]
l Measured flow rate of liquid from the i
discharge tank to the initial dilution stream.
d l
F Flow Rate, Initial Dilution
[cfs]
Measured flow rate of the initial dilution stream which carries the radionuclides to the unrestricted area boundary (e.g.,
circulating cooling water or blowdown from a j
cooling tower or lake).
A.2.4 Tank Overflow l
l Requirement
' O To limit the consequences of tank overflow, standard Technical Specifications (References 2 and 3) limit the quantity of radio-activity which may be stored in unprotected outdoor tanks (see Specification 3.11.1.4).
Unprotected tanks are tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and sur-l rounding area drains connected to the liquid radwaste treatment system.
The specific objective is to provide assurance that in I
the event of an uncontrolled release of a tank's contents, the resulting radioactivity concentrations would be less than the l
limits of 10 CFR 20, Appendix B, Table II, Column 2 a,t the i
nearest potable water supply and the nearest surface water supply in an unrestricted area.
A-42 i
REVISION 0.K JANUARY 1993 A.2.5 Operability and Use of the Liquid Radwaste Treatment System Requirement l
Standard Technical Specifications (References 2 and 3) require that the liquid radwaste treatment system be operable and that appropriate portions be used to reduce releases of radio-activity when projected doses due to the liquid effluent from each reactor unit to unrestricted areas exceed either of the following (see Specification 3.11.1.3):
0.06 mrem to the whole body in a 31 day period.
0.2 mrem to any organ in a 31 day period.
This provision is related to requirements of 10 CFR 50, Appendix I.
1 Section 12.3 of each station's RETS contain a somewhat similar
]
(
l provision.
N Each station except Zion is required to project doses due to liquid releases at least once per 31 days.
Zion is required to project doses due to liquid releases at least once per month.
Equation Offsite doses due to projected releases of radioactive materials in liquid effluents are calculated using Equation A-29.
Projected radionuclide releases are used in place of measured releases Ai.
o A-44
REVISION O.K JANUARY 1993 l
A.2.6 Drinking Water Five stations (Braidwood, Dresden, LaSalle, Quad Cities, and Zion) have requirements for calculation of drinking water dose i
1 that are related to 40 CFR 141, the Environmental Protection Agency National Primary Drinking Water Regulations.
These are
?
discussed in Section A.4.
A.2.7 Non-routine liquid release pathways l
Cases in which normally non-radioactive liquid streams (such as j
the Service Water) are found to contain radioactive material j
are non-routine and will be treated on a case specific basis if and when they occur.
Since each station has sufficent capacity to delay a liquid release for reasonable periods of time, it is.
I expected that planned releases will not take place under these circumstancta.
Therefore, the liquid release setpoint calculations need not and do not contain provisions for
()
treating multiple simultaneous release pathways.
1 A.3 TOTAL DOSE DUE TO THE URANIUM FUEL CYCLE (40 CFR 190)
Requirement Standard Technical Specifications (References 2 and 3) limit the annual (calendar year) dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources to the following (see Specification 3.11.4):
Less than or equal to 25 mrem to the whole body.
e Less than or equal to 25 mrem to any organ except the
]
thyroid.
Less than or equal to 75 mrem to the thyroid.
e O
A-45
,-y, g
-,w-..-m,
REVISION 0.K JANUARY 1993 This provision implements a requirement of 40 CFR 190.
i Each station's RETS contain a somewhat similar provision.
For s_.-
exaxt requirements, see the following sections:
Braidwood 1/2:
12.4 Byron 1/2:
12.4 l
LaSalle 1/2:
12.4 i
Dresden 1:
12.4 Dresden 2/3:
12.3 and 12.4 Quad Cities 1/2:
12.3 and 12.4 Zion 1/2:
12.3 and 12.4 i
o t
'%Y l
O
/
w l
A-45a i
l l
l l
REVISION 0.K JANUARY 1993 When Compliance Assessment is Required.
In both the standard
[ )')
Technical Specifications and the RETS of each station, calculations of total dose are required only when calculated l
l i
I doses from releases exceed certain levels.
In the standard Technical Specifications (References 2 and 3), these levels are twice the following limits:
The RETS limits on dose or dose commitment due to radio-active materials in liquid effluents from each reactor unit (1.5 mrem to the whole body or 5 mrem to any organ j
during any calendar quarter; 3 mrem to the whole body i
or 10 mrem to any organ during any calendar year).
e The RETS limits on air dose in noble gases released in gaseous effluents from each reactor unit (5 mrad for gamma radiation or 10 mrad for beta radiation during any calendar quarter; 10 mrad for gamma radiation or 20 mrad for beta radiation during any calendar year).
j e
The RETS limits on dose due to iodine-131, iodine-133, tritium, and radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each reactor unit (7.5 mrem to any organ during any calendar quarter; 15 mrem to any organ l
(,
during any calendar year).
There are similar criteria for the stations.
See the station RETS for exact requirements.
Comparison with 10 CFR 50, Appendix I.
The differences between the dose limits in 40 CFR 190 and those in 10 CFR 50, Appendix I include the following:
10 CFR 50, Appendix I deals only with radiation due to radioactivity released beyond the boundary of a site.
i A-47 l
l
REVISION 0.K JANUARY 1993 Deff Dose to Organ j Due to Radiation
[ mrem]
j from Gaseous and Liquid Effluents-()
Sum of dose.and dose commitment to i
organ j due to release of radioactive j
materials in gaseous and liquid effluents from the station.
Drad Dose to Organ j due to Noneffluent
[ mrem]
j Sources of Radiation Dose to Organ j due to noneffluent sources of radiation associated with the station (see Section A.3.2).
Doth Dose to Organ j Due to Radiation
[ mrem]
j f*mm Other_ Operations I
Sum of dose and dose commitment to j
organ j of any 27mber of the public in i
the vicinity of ;he station due to i
operations of the uranium fuel cycle other than those of the station.
]
Application I
l l
When dose due to the uranium fuel cycle is required, an initial j
dose estimate is obtained using Equation A-34 as follows:
lO l
e For a member of the public in each sector, the dose (including dose commitment) due to all effluent pathways (airborne and liquid) is obtained for each organ (whole body, skin, thyroid, etc.) (see Section l
A.3.1).
e To this number is added the maximum dose to that organ due to confined sources of radiation (see Section A.3.2) and the dose to that organ in that sector due to other sources of the uranium fuel cycle (see Section A.3.3).
?
l For each target organ, the doses due to all pathways e
l are summed to yield a total dose in each sector, o
For each target organ, the maximum of the 16' sector e
values is determined, and this maximum is compared with j
the 40 CFR 190 limit.
1 If the initial dose estimate exceeds the 40 CFR 190 limit, the assumptions involved are examined in accordance with the O
A-49
I
~
REVISION 0.K JANUARY 1993 guidelines in Reference 46, and, if necessary, the calculation
()
is revised.
A.3.1 Initial Estimate of Dose Due to Effluents from a Station Radiation dose and dose commitment due to radioactivity in gaseous and liquid effluents are evaluated using the equations j
given in Sections A.1 and A.2.
For each receptor (member of public), the sum of dose and dose commitment to each organ is l
i obtained in each sector by summing the contributions from all applicable pathways.
Assumptions for various pathways are as i
described below, i
i I
Whole body dose is obtained in each sector as the sum of the l
l following:
e Whole body dose due to gamma radiation from noble gases l
l in the plume for a receptor located at the site boundary in the sector (see Equation A-6 and Table F-7 i ()
of Appendix F).
)
l e
Whole body dose due to ground deposition for receptor at the point of maximum offsite D/Q in the sector (see l
Equation A-14 and Table F-5 of Appendix F).
e Whole body dose commitment due to inhalation of l
airborne effluents at the point of maximum offsite X/Q in the sector (see Equation A-17 and Table F-5 of l
Appendix F).
e Whole body dose commitment due to ingestion of leafy vegetables and produce from a hypothetical garden located at the point of maximum offsite D/O in the l
sector (see Equation A-18 and Table F-5 of Appendix F).
I e
Whole body dose commitment due to consumption of milk f r o m t h e n e a r e s t m i l k p r o d u c e r i n t h e s e c t o r -( s e e Equation A-18 and Table F-6 of Appendix F).
e Whole body dose commitment due to consumption of meat from the nearest meat producer in the sector (see Equation A-18 and Table F-6 of Appendix F).
e For the Braidwood and Zion sites, whole-body dose O-commitment due to consumption of drinking water from the community water supplies nearest the stations A-50
-.m,
__-._._,,__,,_.m.,y,_,m,
,ye.p_%,..
REVISION 0.K JANUARY 1993 l
(see Equation A-30).
This pathway is not considered I
for the other four nuclear stations because their nearest downstream community water supply locations are so far distant that it is unlikely that a nearest resident to a station would be regularly consuming water containing radioactivity from station effluents, Whole body dose commitment due to consumption ~of fish e
from the "near-field" location (see Equation A-31).
The organ dose is obtained-in each sector as the sum of the following:
1 Organ dose due to the gamma radiation from plume for a e
l receptor located at the site boundary in the sector.
The organ dose is assumed equal to the whole body dose
]
(see Equation A-6 and Table F-7 of Appendix F).
Organ dose due to ground deposition for a receptor at e
the point of maximum offsite D/Q in the sector (see Table F-5 of Appendix F).
The organ dose is assumed equal to the whole body dose (see Equation A-14).
Organ dose commitment due to inhalation of airborne e
l effluents at the point of maximum offsite X/O in the lO sector (see Equation A-17 and Table F-5 of Appendix F).
o Organ dose commitment due to ingestion of leafy vegeta-bles and produce from a hypothetical garden located at l
the point of maximum offsite D/Q in the sector (see Equation A-18 and Table F-5 of Appendix F).
~
Organ dose commitment due to consumption of milk from o
the nearest milk producer in the sector (see Equation i
l A-18 and Table F-6 of Appendix F).
1 I
e Organ dose commitment due to consumption of meat from the nearest meat producer in the sector (see Equation A-18 and Table F-6 of Appendix F).
e For the Zion. site, organ dose commitment due to consump-tion of drinking water (see Equation A-30).,This path-way is not considered for other nuclear stations because their nearest downstream community water supply locations are so far distant that it is unlikely that a nearest resident to a station would be regularly consum-ing water containing radioactivity from station effluents.
Organ dose commitment due to consumption of fish from
()
the "near-field" location (see Equation A-31).
A-51 1
I
REVISION 0.K JANUARY 1993 A.3.2 Dose Due to Contained Sources O
There are presently two " contained" sources of radioactivity
)
which are of concern in offsite radiological dose assessments.
The first source is due to gamma rays from nitrogen-16 carried over to the turbine in BWR steam.
The second source is due to gamma rays associated with radioactive material resident in onsite radwaste storage facilities.
Gamma radiation from these sources contributes to the whole body dose.
A.3.2.1 BWR Skyshine The contained onsite radioactivity source which results in the most significant offsite radiation levels at nuclear power stations is skyshine due to nitrogen-16 (N-16) in turbines and steamlines at boiling water reactors (BWRs).
The N-16 that gives rise to BWR skyshine is produced by neutron activation of O-16 (oxygen-16) in reactor coolant as the coolant passes through the core.
The N-16 travels with steam to the turbine while decaying with a half-life of about 7 j
seconds and producing 6MeV to 7MeV gamma rays.
Typically, offsite dose points are shielded from a direct view of components containing N-16, but there can be skyshine radiation at offsite locations due to scattering of gamma rays off the mass of air above the steamlines and turbine (see Figure A-1).
The offsite dose rate due to skyshine has been found to depend on several factors:
The dose rate decreases as distance from the station increases.
The dose rate increases non-linearly as the level of power being generated increases.
The dose rate increases if a nuclear power station adds hydrogen to its reactor coolant, an action sometimes
[
taken in order tc improve coolant chemistry (see
'(
Reference 39).
I j
A-52
I REVISION 0.K JANUARY 1993 To calculate offsite dose in a given time period due to
()
skyshine, a nuclear power station must keep track of the following:
1 e
The total gross energy Eh produced with hydrogen being added.
e The total gross energy En produced without hydrogen being added.
The turbines at BWR sites are sufficiently close to each other that energy generated by the two units at each site may be summed.
An initial estimate of BWR skyshine dose is calculated per the following equation:
Dsky
+ M E ) x E{ OF SF exp(-0.007R ))
(A-35) j =(K) (E hh k
k k
g This equation applies for each organ j.
The summation is over all locations k occupied by a hypothetical maximally exposed member of the public characterized by the parameters specified s.
in Table F-8.
The parameters in Equation A-35 are defined as follows:
Dsky Dose Due to N-16 Skyshine
[ mrem]
j i
Gamma dose to organ j due to BWR N-16 skyshine for the time period of interest.
The dose is assumed to be the same for all organs and the whole body.
e O
t A-53 1
l l
l REVISION 0.K JANUARY 1993 K
Empirical Constant
[ mrem /(MWe-hr)]
i rN A constant determined by fitting data
(
measured at each station.
See Table F-8 of Appendix F for Dresden, LaSalle and Quad Cities Stations.
E Electrical Energy Generated Without
[MWe-hr]
g Hydrogen Addition Total gross electrical energy generated i
without hydrogen addition in the time period of interest.
Electrical Energy Generated with
[MWe-hr]
Eh Hydrogen Addition Total gross electrical energy generated with hydrogen addition in the time period of interest.
Multiplication Factor for Hydrogen Addition Mh l
Factor by which offsite dose rate due to skyshino is multiplied when there is hydro-gen addition.
Hydrogen addition increases main steam line radiation levels by up to a factor of approximately 5 (see page 8-1 of Reference 39).
Mh is station specific.
l OFk Occupancy Factor
()
The fraction of time that the hypothetical subjeci of the calculation spends at location k in the time period of interest.
See Table F-8 of Appendix F.
SFk Shielding Factor A dimenionless factor which accounts for shielding due to occupancy of structures.
SFk = 0.7 if there is a structure at location k; SFk = 1.0 otherwise.
See Table F-8 of Appendix F.
0.007 Empirical Constant
[m-1]
A constant determined by fitting data measured at the Dresden Station (see Reference 45).
Distance
[m]
Rk Distance from the turbine to location k.
l See Table F-8 of Appendix F.
I
%v A-54 l
l
REVISION 0.K JANUARY 1993 i
A.3.2.2 Radwaste Storage Facility Skyshine o
Buildings to house radioactive waste are in place to hold this i
material prior to shipping to an offsite disposal site.
When these buildings are employed for their intended use, a skyshine dose assessment will be required to complete the 10CFR20 and j
40CFR190 dose assessment.
f t
t i
I L
O
\\
e O
A-54a
REVISION 0.K JANUARY 1993 Air or Water for Occupational Exposure," NBS Handbook 69 as O'
amended August 1963, U.S. Department of Commerce.
If two or more radionuclides are present, the sum of their annual dose equivalent to the total body or to any organ shall not i
exceed 4 millirem / year.
TABLE A -- AVERAGE ANNUAL CONCENTRATIONS ASSUMED TO PRODUCE A TOTAL BODY OR ORGAN DOSE OF 4 MREM /YR pCi per Radionuclide Critical Organ liter i
Tritium Total body 20,000 Strontium-90 Bone marrow 8
A.4.2 Station Requirements Four stations have requirements for calculation of drinking water dose that are related to 40 CFR 141.
The requirements
(}
are as follows:
e Dresden and Quad Cities The surveillance requirement in Section 12.3 requires that doses at the nearest community water system be projected using methods prescribed in the ODCM at least j
once per 92 days.
When the projected annual whole body or internal organ dose at the nearest downstream community water system is equal to or exceeds 2 mrem from all radioactive materials released in liquid effluents from the station, a special report to the operator of the community water system is required.
The report is prepared to assist the operator in meeting the requirements of 40 CFR 141.
LaSalle Action a of Section 12.3 requires a report to the NRC of radiological impact on finished drinking water supplies at the nearest downstream drinking water source whenever the calculated dose from the release of radioactive materials in liquid effluents from either l
unit exceeds the limits, 1.5 mrem.
l A-56 l
l REVISION 0.K JANUARY 1993 i
Table A-1 i
i Release Point Classifications Release Release Point l
Station Point Classificationa i
Braidwood 1 & 2 Vent Stacks Vent (Mixed Mode)
Byron 1 & 2 Vent Stacks Vent (Mixed Mode) l l
Dresden 1 Plant Chimney Stack (Elevated) l Dresden 2 & 3 Chimney Stack (Elevated) i Reactor Building Vent (Mixed Mode)
Ventilation Exhaust Stack 1
l l
LaSalle 1 & 2 Main Station Stack (Elevated)
Vent Stack Standby Gas Stack (Elevated) b Treatment Stack Quad Cities 1 & 2 Chimney Stack (Elevated)
Reactor Building Vent (Mixed Mode)
Ventilation Exhaust Stack Zion 1 & 2 Vent Stacks Ground l
j aThese classifications are based on Sargent & Lundy NSLD Calculation No. CEC-4-88, Rev.
O, 10/19/88.
The defini-tions of release point classifications (stackt vent, and ground level) are given in Section 4.1.4.
bThe LaSalle standby gas treatment stack is located,inside the main station vent stack.
O A-58 i
---a
i REVISION 0.K l
JANUARY 1993 i
Table A-2
(
Nearest Downstream Community Water Systems Characteristics of Nearest Affected Downstream Community Water Supply i
Other CECO Nuclear CECO Nuclear Facilities Location Stations i
Upstream of and Upstream of Station Station Distancea Water Supply Braidwood None Wilmington, None 5 river miles D
Byron None None within NA 115 river milesc Dresden Braidwood
- Peoria, Braidwood l
106 river LaSalle miles LaSalle Braidwood
- Peoria, Braidwood
(~)
Dresden 97 river Dresden miles I
Quad Cities None E. Moline, Nonec 16 river miles Zion None Lake County None i
l
- Intake, j
1.4 miles aTable E-2 of Appendix E provides the bases of the location and distance data.
bNA = not applicable.
For purposes of the calculations of this manual, there are no community water supplies which are considered to be affected by liquid effluents from the Byron Station.
This is based on the absence of community water supplies between the station liquid discharge to the Rock River and the confluence of the Rock and Mississippi Rivers, 115 miles downstream.
(
A-59
t REVISION 0.K JANUARY 1993 Table A-2 (continued)
O-Nearest Downstream Community Water Systems
[
l i
cByron Station discharges its radioactive liquid effluents to the Rock River.
There are no community water supplies on the i'
Rock River downstream of the station discharge.
The Rock River flows into the Mississippi River 115 miles downstream of the station discharge.
The confluence point is below the intake of the E. Moline water supply, which is affected by-discharges from Quad Cities.
l i
L l
l l
l l
O i
[
e s
e e
O A-60
REVISION 0.K JANUARY-1993 APPENDIX D GENERIC DATA l
D.1 INTRODUCTION i
This appendix contains offsite dose assessment data common to one or more of the stations.
D.2 DOSE COMMITMENT FACTORS Dose commitment factors are given in Tables D-1 through D-5 as follows:
Pathway Infant Child Teenacer Adult Inhalation Table D-3 Table D-2 Table D-la Table D-1 Ingestion Table D-5 Table D-4b Table D-4a Table D-4 i
These tables are from Regulatory Guide 1.109 (Reference 6).
Each table provides dose factors for seven organs for each of 73 radionuclides.
For radionuclides not found in these tables, dose factors will be derived from ICRP 2 (Reference 47) or NUREG-0172 (Reference 48).
1 O
D-1 l
w-
+-e-w
--e-+r
- ,,w3-e
-,,m-we-,w_,y-w
l REVISION 0.K JANUARY 1993 l
Table D-la Inhalation Dose Factors for Teenager (mrem per pCi Inhaled) i l
i i
l
.NuCLICE SONE LtvEA T.s00Y THvtot0 E!DNEY LUNC Cl-LLI l
a 3
NO Data 1.59E-CF 1.59E.07 1.59E-07 1.59F-07 1.59E-C7 1.59E-07 C 16 3.2bE-C6 6.L9E-07 6.09E-07 6.09E-07 6.09E-37 6.09E-77 6.09E-07 na 26 1.72E-06 1.72E.06 1.7JE-06 1.72E-06 1.72E-06 1.72E-06 1.72E-06 I
- 12 2.36E-04 1.17E-03
- 8. 95 E -0 6 NO DATA NC OATA No cafa 1.16F-05 CR St NO Data No cara 1.6?F-38 9.27E-09 1.942-e9 2 62E-06 3.75E-07 MN 54 NO Cafa 6.31F-06 1.03E-06 NO DATA 1.59E-n6 2.48E-06 8.35E-C6 l
i
- N 56 40 DATA 2.12 -10 3.15E-11 NU DATA 2.24t-in 1.90E-C6 7.18E-06 FE 55 4 18E-C6 2.98t-06 6.93E-3F NO DATA NO Data 1.55E-05 7.99E-07 F E 51 1 19E-C6 6.62i-06 1.71E-06 40 DATA NO DATA 1.91E-04 2.23E-C5 CC 58 NO Data 2.39f-C7 5.47E-37 NO DAfa ha CAfa 1.68E-04 1.19E-05 CD 00 10 CAta 1.812.C6 2 48s-06 NO 04T4 NC DATA 1.09E-03 3.2 4 E-0 5 NI 63 7.25E-C5 5.*3E-06 2.6FE-06 NO Data NC Data 3.84E-05 1.77E-04 l
41 65 2.73E-19 3.o65-11
- 1. 59 E -1 1 40 DATA NO Data 1.17E-06 4.59f-06 CO 64 NO DATA 2.54t-10 1 06E-1C NO DAT4 8.01E-10 1.19E-06 7.68E-04 IN 65 4.E2E-C6 1.o7E-05
- 7. 80 E-0 6 40 Data 1.08E-9S 1.55E-04 5.83E-06 f
IN 69 6.04E-12 1.15F-11 8.07E-13 NO DATA 7.53E-12 1.98E-07 3.16T-08 t
8a 8J NO Data 90 3Afa 4.30E-08 NO Data NC Data NO Data LT E-24
\\~
34 84 NO Data eu caTa 5.61E-08 NO DATA NC Data NO Dafa LT E-24 l
RR 85 NO Data 40 Cafa 2.29E.09 40 DAfa NO DATA NO DATA LT E-24 A8 86 NO Daft 2.381-05 1.05E-05 NO Data NC DATA NO DATA 2.21E-06 a9 88 40 DATA 6.82E-08 3.60E-08 NO Dafa NO DATA NO DATA 3.65E-15 88 89 NO Data 4.6CE-08 2.91E-08 NO Data NC DafA NO DATA 4.22E-17
%R 89 5.43E-C5 No Data
- 1. 56 k -0 6 NO Data NC DATA 3.02E-C4 4.64E-05 SR 90 1.35E-02 NO Daft
- 8. 3 5 E -0 4 NO Daft NC Data
- 2. 04 E-0 3 9.56E-05 54 91 1.10E-08 NO CATA 4.39E-10 NO DaT2 NO Data 7.59E-06 3.24E-05 SR 92 1.19E-C9 NO CATA 5.08E-11 No DAfa NO Data 3.43E-06 1.49E-05 Y 90 3.73E-07 NO DaT4
- 1. 00 E -0 8 No DATA NO Data 3.66E-05 6.19E-05 Y 91 4.63E-11 NC DATA 1.77E-12 40 Data NO DATA 4.00E-07 3.77E-09 Y 91 8 26E-05 NO 04TA 2.21E-36 NO DATA NO DATA 3.67E-04 5.11E-05 Y
94, 1.86E-09 NO DATA
- 5. 36 E -1 1 NO Dafa NO DATA 3.35E-06 2.06E-05 l
OLJ D-4a
REVISION 0.K JANUARY 1993 l
Table D-la (cont'd.)
Inhalation Dose Factors for Teenager l
(mrem per pCi Inhaled) l l
NUELICE SONE L i vF 4 T. 40DT THYR 010 KIONEY LuwC GI-LL1 Y 93 1 69E-08 NO Dsfa
- 4. 6 5 E-10 40 DATA 40 DATA 1.04E.05 7.24E-05 2R 95 1.82E-05 5.73E-06
- 3. 96 E -0 6 NO DATA 8.42E-06 3.36E-04 1.86E-05 2R 17 1.72E-08 3.40E-09
- 1. 5 7 E-0 9 40 DATA 5.15E-09 1.62E-05 7.88E-05 l
NB 9h 2.32E-06 1.29E-06
- 7. 0s E-0 7 N0 DATA 1.25E-06 9.39E-05 1.21E-05 mo 99 740 DATA 2.11E-CB
- 4. 0 lE -0 9 MD CATA 5.14t-08 1.92E-c5 3.36E-05 ft 998 1.FIE-13 4.8 3E-I l 6.24E-12 90 DATA 7.20E-12 1.44E-07 7.66E-07 TC101 7.40E-15 1.0SE-14 1.03E-13 NO DATA 1.50E-11 8.34E-00 1.09E-16 RU103 2.63E-07 40 CATA 1.12 F-0 7 NO DATA 9.29E-07 9.79E-05 1.36E-05 9u105 1.40E-10 40 DATA 5.42E-11 NO DATA 1.16E-10 2.27E-06 1.13E-05 EU106 1.23E-05 NO DATA
- 1. 5 5 E -0 6 40 DATA
- 2. 30 E -05
- 2. 01 E-0 3 1.20E-04 AG110p 1.73E-06 1.64E-06
- 9. 99E -0 7
.NO DATA 3.13E-06 8.64E-04 3.418-05 TE1248 6.10E-07 2.80E-0F 8.34E-00 1.75E-07 40 DATA 6.7SE-05 9.38t-06 TE127m 2.25E-06 1 02E-06
- 2. 7 3E-0 7 5.48E-07 8.17E-06 2.07E-04 1.998-05 TE127 2.51E-10 1.14F-10 5.52E-11 1.77E-10 9.10E-10 1.40E-06 1.01E-05 TE129#
1.74E-06 8.23E-07 2.81E-07 5.72E-07 6.49E-06 2.67E-04 5.06E-05 TE129 8.87E-12 4.22E-I2 2.20E-12 6.4AE-12 3.32E-11 4.12E-07 2.02E-07 TE131*
1.23E-08 7.51E-09 5.0 3 E-0 9 9.04E-09 5 49E-08 2.97E-05 7.76E-05 TE131 1.97E-12 1.04E-12 6.30E-13 1.55E-12 7.72E-12 2.92E-07 1.09E-09 TE132 4.50E-08 1.e3E-08 2.74E-Os 3.07E.08 2.44E-07 5.61E-05 5.79E-05 l
I 130 7.80E-07 2.24E-06 S.96E-0 7 1.86E-04 3.44E-06 40 DATA 1.16E-06
! 131 4.43E-06 6.14E-06
- 3. 30E-0 6 1.83E-03 1.05E-05 m0 DATA 8.11F-07 l
I 132 1.99E-07 5.47E-07
- 1. 9 7 E-0 7 1.89E-05 8.65E-07 ho DATA 1.5 9E-07 1 133 1.52E-06 2.36E-06
- 7. 70E-0 7 3.65E-04 4.49E-06 NO DATA 1.29E-06 1 134 1.11E-07 2.90E-07
- 1. 05 E-0 7 4.94E-06 4.58E-07 b0 DATA 2 55E-09 I 135 4.62E-07 1.18E-06
- 4. 36 E -0 7 7.76E-05 1 86E-06 40 DATA 8.69E-07 C5134 6.20E-05 1.41E-04
- 6. 86E-0 5 40 DATA 4.69E-05 1.83E-05 1.22E-06 C1116 6.44E-06 2.42i-05
- 1. 71 E-0 5 40 DATA 1.38E-05 2.22E-06 1 16E-06 C5137 8.38E-05 1.06E-04
- 3. 8 9 E-0 5 40 DATA 3.80E-05 1 51E-05 1.06E-06 C5136 5.82E-08 1.07E-07 5.5 e E-0 8 20 DATA 8.28E-08 9.86E-09 3.38E-11 Salter 1.67E-10 1.18E-13 4.07E-12 40 DATA 1.11E-13.8.08E-07 8.06E-07 l
l D-4b
REVISION 0.K JANUARY 1993 Table D-la (cont'd.)
Inhalation Dose F dors for Teenager (mrem per PCi Inhaled)
NUCLl?E GONE L i vE R T.3C cv TMveOlc SIDNEY LUNG GI-LLI SA160 6.86E-06 4.38F-09 6.40E-37 NO D&fa 2.85E-69 2.54E-04 2.86E-05 Da161 1.TBE-11 1.22F-16 S.9sE-13 NO DATA 1.23E-14 4.11E-07 9.33E-14 sa162 6.62E-12 6.6JE-1S 2.86E-13 NO Cafa 3.92E-15 2.39E-07 5.99E-20 LA160 2.99E-08 2.95 -08 7.82[-09 40 Cara 40 DATA 2.68E-05 6.09E-05 L4162 1.20E-10 5.J12-11 1.32E-11 NO Data 40 DATA 1 27E-06 1.50E-06 CE161 3.55E-06 2.JF0-06
- 2. 71 E-0 7 NO DAT4 1 11E-06 f.67E-OS 1.58E-05 i
CE143 3.32E-08 2.42E-C8
- 2. 70 E-0 9 10 DATA 1.086-08 1.63E-09 3.19E-05 CE144 6.11E-04 2.53E-06
- 3. 2 4 E -0 5 %e OAfA 1.51E-04 1.6TE-03 1.08E-04 l
P'163 1.67E-06 6.66E-C7 8.28E-08 40 nafa 3.86E-07
- 6. 04 E-P 5 2.67E-05 valsa 5.J7E.12 2.2C5-12 2.72E-13 NO DATA 1.26E-12 2.19E-C7 2.94E-16 WC1 7 9.81E-C7 1.07E-06
- 6. 41 E -0 8 40 DATA 6.25E-C7 4.65E-05 2.28F-05 w 187 1.50E-01 1.228-C9 6.21E-1C NO cara NC OATA 5.92E-06 2.21E-05 l
l NP239 4.23E-06 3.v9E-01
- 2. 21 E-0 9 40 DATA 1.25E-C8 A.118-06 1.65E-05 1
I l
l i
s l
I i
l l
l 1
i i
D-4c
REVISION 0.K JANUARY 1993 Table D-4a Ingestion Dose Factors for Teenager (mrem per pCi Ingested)
NUCLICE BONE LIvfa T.40CY THYR 010 KIDNEY LUNG GI.LLI H
I NO DAfa 1.065-C7 1.coE-07
- 1. 06r-0 7 1.06E.07 1.066-07 1 06E-07 L 16 4.06E.06 E.12F-07 8.12 E.0 7 8.12F-07 8.12E-07 P.12E.07 8 12 F-0 7 na 26 2.30E.06 2.20E-96
- 2. 3 C E -0 6
- 2. 30E-06 2.30t-06 2.10E-36 2.30E.06
- 32 2.76E 04 1.71C-05 1.07E-05 NO DATA NC DATA NO DATA 2.32E.05 Ca 51 Nu DATA NO Data 3 60E.01 2.00E.09 7.89t.10 5.14E-09 6.05E.07
- N 54 NO DATA 5.90E.00 1 17E.06 NO DATA 1.76E.06 NO DAfa 1.21E.05 mN 56 40 DATA 1.30E-07 2.4tE.08 NO DATA 2.00E.07 NO DATA.
1.04E 05 FF S3 3.78E.06 2.68F-06 6.25E-07 40 DAfa NC Data 1.70E.36 1 16E.06 FE 59 5.87E-06 1.J7E.05 5.29E-06 40 DATA NO DATA 4.12E-06 1.24E.05 CO 58 NO DATA, 0.726 07
- 2. 24 E.0 6 Nu DATA NO DATA NO Data 1.34E-05 CU 60 40 DATA 2.CIE.S6
- 6. 3 3E.0 6 NO DATA NO DATA NO DATA 3.66E-05 l
nt 63 1.77E-04 1.250 0%
o.0DE-06 90 DATA No DATA NO DATA 1 99E-06 NI 65 7.49E-07 9.37E-04 4.3eE 08 40 DATA NO DATA No DATA 5.19E.06 CU 64 NO DATA 1.15E 07 5.41E.08 NU DATA 2.91E 07 NO DATA S.92E.06 ZN 65 5.76E-06 7.COE.C5
- 9. 3 s E-0 6 NO DAT4 1.7BE-95 NO DATA 8.47E.06
/"
tw 67 1 47E 08 2.60s-08 1 96E.01 NO DATA 1.a1E.08 No Data 5.16E-08 BR 84 NO DATA NO DATA S. 74 E -0 8 NO DATA NO DATA NO DATA LT E-24 EA 86 NO DATA NO CATA
- 7. 2 2 E -0 8 NO DAT4 NC DATA NO DATA LT E-24 I
6A 8%
NO DATA NO DATA 3.05 E -0 9 NO DATA 40 DATA NO DATA LT E.24 an 86 No DATA 2.18E.05 1 40E-05 40 DATA NO DATA NO CATA 4.41E-06 R8 88 10 DATA E.52E.08 4.56E-08 NO DATA NO DATA NO D&TA 7.30E.15 R8 89 NO DATA 5.30E-08 3 89E.08 NO DATA NO DATA NO DAfa 8.43F-17 54 81 4.40E.04 NO DATA 1.26E.05 NO DATA 40 O&fa NO Daft 5.24E-05 I
$4 90 8.30E 03 NO DATA
- 2. C S E -0 3 40 DATA 40 DATA MO DATA 2.135-C4 SR 11 8.C)E-06 40 DATA 3 21E.07 No DATA NO DATA NO DATA 3.66E.05 54 92 3.05E.06 NO DATA 1 30E 07 NO DATA NO DATA NO DATA 7.77E 05 Y 90 1.37E.08 NO DATA 3 61E.10 NO Data NO DATA NO DATA 1.13E.04 Y 9tp 1 29E-10 NO DATA 4.93E.12 NO DATA NC DATA NO DATA 6.09E-09 y 91 2 01E.07 40 DATA
- 5. 31E -0 9 NO DATA NO DATA NO DATA 8.24E-05 i
l V 92
_E.21E.09 NO DATA 3.5GE.11 NO DATA 40 DATA 40 OATA
- 3. 32 E-0 5 i
l l
D-13a r
REVISION 0.K JANUARY 1993 Table D-4a (cont'd.)
Ingestion Dose Factors for Teenager (mrem per pCi Ingested)
NUCL10E 60NE LIVER T.4cov THYR 0fD KIDNEY LUNC Cl-LLI v 93 3.81E-09 40 CATA 1.05E-10 NO D A T'A 40 CATA 40 DATA 1.17E-04 Zu 93 4.12E-0M 1.30C-08 8.9*E-09 40 DATA 1.916.en 40 CATA 1.00E-05 ft 97 2.37E-09 4.09E-10 2.16E-10 90 DATA 7.11E-10 No CATA 1.27E-04 sp 95 8.22E-09 4.>ei-09
- 2. 31 E -0 4 NO DATA 4.42E-09 40 DATA 1.95E-05
- 0 99 90 CATA 6.03E-06 1.13 F -0 6 NO DATA 1.30E-05 40 DATA 1.00E-OS TC 99m 3.12E-10 9.265-10 1.20E-08 NO DATA 1.38E-08 S.14E-10 6.08E-07 TC101 3.60E-10 1.12i-10
- 5. 01 E-0 9 No DATA 9.26E-09 3.12E-10 8.75E-17 RU103 2.55E-07 40 DATA
- 1. 09 E -0 7 NO DATA 8.99E-07 No DATA 2.13E-OS dut05 2.18E-04 NO DATA
- 5. 4 6 E -0 9 NO DATA 2.75E-07 40 DATA 1.76E-05 i
tulC6 3.92E-06 NO DATA 4.946-07 NO DATA T.56E-06 NO DATA 1.88E-04 l
4G110m 2.05E-07 1.94E-07 1.18 E-0 7 40 CATA
- 3. 70E *07 NO DATA 5.45E-05 TE12S* 3 83E-06 1.386-06 S.12 F -0 7
- 1. 0 7E-04 40 DATA M0 DATA 1.13E-DS TF127m 9.67E-06 3.*sE-06 1.1 S E-0 6 2.30E-06 1.92E-05 40 DATA 2.41E-05 IE127 1.58E-07 5.60E-08 3.40 E-0 8 1.09F-07 6.40E-07 NO DAT4 1.22E-OS TE129m 1.63E-05 6.05E-06
- 2. 5 8 E-0 6 S.26E-06. 6.42E-05 20 DATA 6.12E-05 4....
TE129 4.48E-09 1.67F-09 1.09E-08 3.20E-08 1.88E-07 40 DATA 2.4SF-07 TE131*
2.44E-06 1.17E-06 9.76E-07 1.76E-06 1.22E-05 N0 DATA 9 39E-05 TE131 2.79E-08 1 15E-08 8.72E-01 2.15E-08 1.22E-07 m0 DATA 2.29C-09
...... ~.. _........
TE132 3.49E-06 2.21F-06 2.04E-06 2.33E-04 2.12E-OS 40 DATA 7.00E-05 1 130 1.03E-06 2.98E-06 1.19E-06 2.435- 04 4.59E-06 NO DATA 2.29E-06 l
1 1 41 5.8SE-C6 8.19C-06
- 4. 4 0E-0 6 2.39E-03 1.41E-05 40 DATA 1.62E-06 i
l l
t 132 2.79E-07 7.30E-07
- 2. 62 E-0 7 2.46E-05 1.15E-06 NO DATA 3.18E-07 1 133 2.01E-06 3.41t-06
- 1. 04 E -0 6 4.76E-04 S.98E-04 40 DATA 2.58E-06 1 13 1 46E-07 3.87E-07
- 1. 39E-0 7 6.4SE-06 6.10E -0 7 40 DATA S.10E-09 l
~
I 135 6.10E-07 1.57E-06
- 5. 82 E -0 7 1.01E-04
- 2. 48 E-06 NO DATA 1.74E-06 C5134 8.37E-OS 1.97E-04 9.14 E-0 5 40 CATA 6.26E-05 2.39E-05 2.45E-06 C5136 8.59E-06 3.J8E-05 2.2 7 E -0 5 No DATA 1.04E-OS
- 2. 90E-06 2.72E-06 C5137 1.12E-04 1.49E-04 S.19E -0 5 40 DATA S.0 7 E -OS 1.97E-05 2.12E-06 C5138 7.76E-C8 1.49E-07
- 7. 4 5E-0 8 20 DATA 1.10E-07 1.28E-08 6.76E-11 i
l BA139 1 39E-07 9.78E-11
- 4. 0SE-0 9 No DATA 9.22E-11 4.76E-11 1.24E-06
! \\
D-13b
REVISION 0.K JANUARY 1993 i
Table D-4a (cont'd.)
Ingestion Dose Factors for Teenager l
(mrem per pCi Ingested)
WUCL1LE BONE Liven f.90tv inva010 K10NEY LUNG G1.LL1 BA160 2.86E.05 3.44E-09 1 84E-06 NO Jafa 1.18t-09 2.16E-08 4.38E-05 malet 6.71E-04 5.010 11 2.24E-09 Nn Daf a 4.65E-11 5.6sE.11 1.63E-13 SA142 2.99E-C8 2.19E-11 1.86E-09 40 D A T A 2.53t-11 1.99E-11 9.18E-20 LA160 3.68E-09 1.71E-C9 4.55E-10 40 DATA NC DATA 40 DATA 9.82E.i i
i LA162 1.79E.10 7.95E.11 1.98E-11 NO DATA NO DATA ho DATA 2.42E-(
Ch141 1.13E.C8 8.88E.09 1.02E-09 NO DATA 4.18E.09 NO DATA 2.54E-(
Cele) 2.35E-09 1.71L-C6
- 1. 91 E -10 NO DATA 7.67E-10 NO DATA S.14E.s CF166 6.96E 07 2.88E 07 3.74E-08 40 DATA 1.72E-07 NO DATA 1.75F-Os PR163 1.31E.08
- 5. 2 3E.09
- 6. 5 2 E.10 NO DATA 3.64E.09 40 DATA
- 4. t 1 E -e '.
PR146 4.lCE.11 1.76E.11 2.18E-12 NO DATA 1.01E-11 40 DATA 4.f4E-16 i
ND14F 1.l8E-09 1 02F-08 6.11E.10 40 DATA 5.99E.C9 40 DATA 3.68E.05 e 187 1.46E-07 1.19E.07 6.1FE-08 40 DATA 40 DATA NO DATA 1 22E-05 s........................................................................
N#234 1.76E-09 1.66E-10 9.22E-11 NO Data 5.21E-10 40 DATA 2.67E.05 O
I D-13c l
REVISION 0.K JANUARY 1993 Table D-6 l
Miscellaneous Dose Assessment Factors -
Environmental Parameters Parameter and Value Basisa f
= 0.76 7
p fy
= 1.0 d
th
= 0 for pasture grass (milk and meat pathways)
B th
= 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (1 day for leafy vegetables)
B l
th
= 1440 hr (60 days for produce or leafy vegetables)
B th
= 2160 hr for stored feed (milk and meat pathways)
B t
= 720 hr (30 days for milk and meat)
B e
t
= 1440 hr (60 days for produce or leafy vegetables)
B e
ff
= 1.0 May-October C
ff
= 0.0 November-April C
l l
f
= 0.5 C
g A
= 0.0021 hr 1 B
w 2
Y
= 2.0 kg/m for leafy vegetables and produce pathways B
y 2
Y
= 0.7 kg/m for milk and meat pathways B
y t
= 480 hr (20 days)
B s
r
= 1.0 (iodines)
B l
= 0.2 (others)
B Wf
= 50 kg/ day D
tM
= 48 hr (2 days)
B l
i tb
= 175,200 hr (20 years)
E f
= 1.0 May-October C
a f
= 0.0 November-April C
a O
D-17 l
REVISION 0.K JANUARY 1993 Table D-6 (Cont'd) i l
l aBasis key:
B:
Reference 6, Table E-15.
C:
Typical for climate of Illinois and vicinity.
D:
Reference 6, Table E-3.
E:
The parameter tb is taken as the midpoint of plant l
operating life (per Reference 6, Appendix C, Section 1).
l l
s I
D-18 r
l l
..n.
l REVISION 0.K l
JANUARY 1993 i
l Table D-7 Miscellaneous Dose Assessment Factors Consumption Rate Parameters Tvoe Variable Infant Child Teenacer Adult 3
Air Ra(m /yr) 140 3700 8000 8000 M
Milk U (L/yr) 330 330 400 310 a
P(Kg/yr) 0 520 630 520 Produce Ua V(Kg/yr) 0 26 42 64 Leafy Ua Vecetables F
Meat U (Kg/yr) 0 41 65 110 a
W l(
)
Water U (L/hr) 0.038 0.058 0.058 0.083 a
f Fish U (Kg/hr) 0 7.9E-4 1.8E-3 2.4E-3 a
l l
l Reference 6, Table E-5.
l D-19 l
l l
l
REVISION 0.K JANUARY 1993 This page initially left blank.
1 4
l 1
l i
I l
i i
{
l l
I D-20 l
I e
REVISION 0.K JANUARY 1993 Table D-13
! (\\m/~]
Bioaccumulation Factors (B ) to be Used i
in the Absence of Site-Specific Data l
Bi for Freshwater Fish Element (DCi/ko per DCi/L)
H 9.0E-01 C
4.6E+03 Na 1.0E+02 b
P 3.0E+03 Cr 2.0E+02 Mn 4.0E+02 l
Fe 1.0E+02 Co 5.0E+01 Ni 1.0E+02 Cu 5.0E+01 Zn 2.0E+03 Br 4.2E+02
[~
Rb 2.0E+03 Sr 3.0E+01 Y
2.5E+01 1
Zr 3.3E+00 Nb 3.0E+04 l
Mo 1.0E+01 l
Tc 1.5E+01 i
Ru 1.0E+01 Rh 1.0E+01 Te 4.0E+02 s
I 1.5E+01 Cs 2.0E+03 Ba 4.0E+00 La 2.5E+01 l
Ce 1.0E+00 b = Refer to Reference 82.
O D-32 j
REVISION O.K JANUARY 1993 l
Interim Radwaste Storage Facilities lO Interim Radwaste Storage facilities (IRSF) have been constructed at four stations - Dresden, LaSalle, Quad Cities and Zion.
l These facilities were designed to serve as temporary repositories of solidified radwaste before shipment offsite.
The surface dose rate of these containers may be as high as 15 l
R/hr.
l l
l Consideration is also being given to store containers of l
compacted dry active waste (DAW) in Sea-Land containers at all l
nuclear power plant sites.
These may have surface dose rates as l
high as 8 mR/hr at a distance of 2-meters from the container surface.
l l
Both the IRSF and DAW will contribute direct radiation to points in the controlled and unrestricted areas.
Thus a dose i
assessment is required to assure site compliance to the
()
regulations of 40 CRF 190.
The dose due to IRSF's have been calculated in References 56a, b, e and d.
In these calculations, the containers were assumed to have a contact dose rate of 15 R/hr; consideration was given to accessible sites outside of the restricted area boundary, but near the IRSF.
Although some of these sites are less than 200 meters from the IRSF, the annual doses are less than 10% of the 40 CFR 190 limit of 25 mrem / year when realistic occupancy factors are considered.
i l
The above calculations are, of course, estimates as t,he inventories, nuclide mixes, decay times, container self-shielding, and other factors affect the corresponding out-of-building dose rate.
As the IRSF's become operational, the above estimatos will be re-evaluated.
A correlation of in-IRSF dose rate (area radiation monitor reading) with measured I()
out-of-IRSF will be evalur.ted as a better means of quantifying the IRSF offsite dose rate.
E-8 1
REVISION O.K l
l JANUARY 1993 Interim Radwaste Storage Facilities (cont'd.)
O The dose due to storage of Dry Active Waste (DAW) on site in arrays of Sea / Land Vans has been evaluated.
For a design basis source of 8 mR/hr at a distance of 2 meters, calculations (References 57a, b, c and d) show that a dose rate of 1 mrem per year will not be exceeded at the restricted area boundary for realistic combinations of DAW locations and occupancy factors.
Since occupancy at the points of maximum offsite exposure is likely to be much less than 100%, doses due to the interim radwaste storage facilities are judged negligible in comparison with 40 CFR 190 limits.
E.4 BASES OF CHAPTER 4, INTRODUCTION TO METHODOLOGY Most of the material in this chapter is based on Appendix A.
Additional information on bases is provided below.
O E.4.1 Introduction of Time Factors In explaining the equations of Appendix A, a factor t (represen-ting the time period of concern) is sometimes added to the b
e E-Ba O
. - ~..
REVISION O.K JANUARY 1993 in Equation A-17 of Appendix A are identical to the inhalation
()
dose factors provided in Tables E-7, E-9, and E-10 of Regulatory Guide 1.109.
E.9.17 Food Pathways Doses (Section A.l.4.3 of Appendix A and Section B.10 of Appendix B)
The dose commitment due to food pathways is calculated by Equations A-18 through A-27 of Appendix A.
These equations are discussed in Section 4.2.4.
They are like Equations 14 and C-5 through C-13 of Regulatory Guide 1.109 except as follows:
The treatment here neglects the pathway of uptake of radionuclides from soil by edible vegetation (the Biy term in Equation C-5).
The reasons are discussed in Section E.3.1.
l In this manual the equations provided for calculating food pathways dose commitments due to ingestion of carbon-14 and tritium are the same as the equations for dose commitments due to particulate and iodine radio-nuclides.
In contrast, Regulatory Guide 1.109 provides l (
special equations for carbon-14 and tritium (Equations C-8 and C-9).
The decision not to use the Regulatory Guide 1.109 equations was based on the judgment that the added complexity of using the special equations was not justified by the small dose commitments expected due to carbon-14 and tritium.
(Note:
At present, no station i
calculates doses due to carbon-14 releases.
See Section E.4.5 for the reasons.)
e Regylatory Guide 1.109 states, "For radiciodines, the model considers only the elemental fraction of the effluent.
Tha deposition should be computed only for I
that fraction of the effluent that is estimated to be elemental iodine.
Measurements at operating facilities indicate that about half the radiciodine emissions may be considered nonelemental."
Using this rationale, RG 1.109 then halves the deposition rate equation for radioiodines entering the food pathways.
Thi-s ODCM, however, does not include this one-half factor, and thus is conservative be a factor of 2 for the radiciodine food pathway doses.
l l
The dose calculations for particulates and radioiodines account for doses resulting from dry deposition of radioactive materials E-26
REVISION O.K JANUARY 1993 In both models, the decay time between releases of radioactivity and its consumption in fish is taken as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This is the value recommended for the maximally exposed individual in Table E-15 of Regulatory Guide 1.109 (see parameter tp).
E.10.2 Concentration Due to Tank Discharges (Section A.2.3 of Appendix A)
The concentration of radioactivity in tank discharges is calculated by Equation A-33 of Appendix A.
The basis of this equation is explained in Section C.2 of Appendix C.
E.ll BASES OF CALCULATIONS OF TOTAL DOSE FOR THE URANIUM FUEL CYCLE (SECTION A'.3)
]
E.ll.1 Initial Estimate of Dose Due to Contained Sources at a Station (Section A.3.2 of Appendix A)
Annual radiation doses due to contained sources of radioactivity at the stations are judged to be negligible in comparison with
{}
applicable limits except for doses due to BWR turbine skyshine.
This judgement is based on the considerations discussed in Section E.3.3.
l l
I The dose due to N-16 skyshine is calculated by Equation A-35 of Appendix A.
This equation is based on the following:
e Measurements of dose rate due to skyshine made at Dresden, Quad Cities and LaSalle.
An empirical fit to the above data (References 45a, b and c).
Measurements of the radiological effects of h,ydrogen addition to primary coolant at Dresden 2 (Reference 73).
t O
E-29
)
l
M REVISION O.K JANUARY 1993 e
Guidelines for BWR hydrogen water chemistry installa-tions prepared by the Hydrogen Installation Subcom-l,,}
mittee of the BNR Owners Group for Intergranular
\\-
Stress Corrosion Cracking (Reference 39).
References 45a, b and c provide a mathematical expression for calculating an upper bound to skyshine dose when there is no hydrogen addition to primary coolant.
When there is hydrogen addition, the dose is multiplied by a factor of 5.
The value of this factor is based on data and guidelines in References 73 (see Page 4-13) and 39 (see Page 8-1).
Because of natural background radiation, it was only possible to measure skyshine dose rate only to about 600 meters from the turbines.
Beyond this distance, skyshine dose rate was so small that it was masked by fluctuations in the background radiation level (see References 45a, b and c).
Despite this, Equation A-35 of Appendix A is put forth here for use at larger distances.
This is done because estimates of skyshine dose at distances above 600 meters are sometimes needed and because s
(_)
Equation A-35 of Appendix A is consistent with measurements at lower distances.
E.ll.2 Initial Estimate of Dose Due to Other Facilities of the Uranium Fuel Cycle (Section A.3.3 of Appendix A)
In evaluating compliance with 40 CFR 190, radiation doses from other uranium fuel cycle facilities are treated as negligible l
l p
l L/
E-30