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| issue date = 07/17/1992
| issue date = 07/17/1992
| title = Rev 8 to Salem Nuclear Generating Station Odcm.
| title = Rev 8 to Salem Nuclear Generating Station Odcm.
| author name = PALUZZI V
| author name = Paluzzi V
| author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY
| author affiliation = PUBLIC SERVICE ELECTRIC & GAS CO. OF NEW JERSEY
| addressee name =  
| addressee name =  
Line 17: Line 17:


=Text=
=Text=
{{#Wiki_filter:'*.**, ... .. * ;_' SALEM NUCLEAR GENERA'rING'  
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'.-.. :,. -:: OFFSITE O()SE CALCULAT.!'ON M$'UAL , : .. *' *: *:;* * . .; . .;_
SALEM NUCLEAR GENERA'rING' *sT.A'.ttlION
Revision a July.  
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OFFSITE O()SE CALCULAT.!'ON M$'UAL ,                                                       :
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*-------9303090326 930226 PDR ADOCK 05000272 R PDR .*._. .. *-.r.:.. . * ...* *' . . . * *.** *.*;-* ...   
Revision     a July. *-1~9-~
,. I , Salem ODCM 8 SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents Introduction . . . . -.
                                                                .. .   ~ :. *.
-. . * . . . . .
Approval S\)RC   Chaimn:~~~*i                    Datil:             24u~i                                            Mtq. #' *if,/,'-a.Al'
* *
                                                                                          .. *-.r.:..       .   *.*;- *...
* _ * . . . . . . . . 1 1.0 Liquid Effluents Lt Radiation Monitoring Instrumentation
J~ *...**' .
*and Controls . 2 1.2 Liquid Effluent Monitor Setpoint Determination 3 1.2.1 Liquid Effluent Monitors (Radwaste, steam Generator Blowdown and Service Water) * * . 4 1.2.2 Conservative Default Values * * * * * * * . s 1.3 L1quid Effluent Concentration Limits -10 CFR 20 7 1.4 Liquid Effluent Dose Calculations
      ,- -------~-- *--- ----
-10 CFR 50 . * . a 1.4.1 Member of the Public Dose -Liquid Effluents 8 1.4.2 Simplified Liquid Effluent Dose Calculation 10 1.5 Secondary Side Radioactive Liquid Effluents
9303090326 930226 PDR ADOCK 05000272 R                PDR
-Dose calculations During Primary to Secondary Leakage. . 11 1.6 Liquid Effluent Dose Projection
******.*.
13 2.0 Gaseous Effluents 3.0 2.1 Radiation Monitoring Instrumentation and Controls . 15 2.2 Gaseous Effluent Monitor Setpoint Determination
.. 17 2.2.1 Containment and Plant Monitor ... 17 2.2.2 Conservative Default Values *****... 19 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations
-10 CFR 20. * * * * * . 20 2.3.1 Site Boundary Dose Rate -Noble Gases. . 20 2. 3. 2 Site Boundary Dose Rate -., Radioiodine and Particulates
* * * * * :i . . 21 2.4 Noble Gas Effluent Dose Calculations
-10 c::FRt,50
.. 24 2.4.1 UNRESTRICTED AREA Dose -Noble Gases **.. 24 2.4.2 Simplified Dose Calculation for Noble Gases. 25 2.5 Radioiodine and Particulate Dose Calculations
-10 CFR 50. * * * * * * * * * * * * * * * * * . . . 26 2.5.1 UNRESTRICTED AREA Dose -Radioiodine and Particulates
* * * * * * . . 26 2.s.2 Simplified Dose Calculation for and Particulates
* * * * * * * * * * . . 27 -2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations
* * * * * * * * * * * * . . . 28 2.7 Gaseous Effluent Dose Projection
*******..
32 Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY.


===3.2 Doses===
Salem ODCM    Rev~ 8 SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents Introduction . . . . -. - . . * . . . . . * *
to MEMBERS OF THE PUBLIC -40 CFR 190
* _ * ........          1 1.0  Liquid Effluents Lt Radiation Monitoring Instrumentation *and Controls . 2 1.2 Liquid Effluent Monitor Setpoint Determination              3 1.2.1 Liquid Effluent Monitors (Radwaste, steam Generator Blowdown and Service Water) * * . 4 1.2.2 Conservative Default Values * * * * * * * . s 1.3 L1quid Effluent Concentration Limits - 10 CFR 20            7 1.4 Liquid Effluent Dose Calculations - 10 CFR 50 . * . a 1.4.1 Member of the Public Dose - Liquid Effluents 8 1.4.2 Simplified Liquid Effluent Dose Calculation 10 1.5 Secondary Side Radioactive Liquid Effluents - Dose calculations During Primary to Secondary Leakage. . 11 1.6 Liquid Effluent Dose Projection        * * * * * * . * . 13 I ,
* 3. 2 .1 Effluent Dose Calculations  
2.0  Gaseous Effluents 2.1 Radiation Monitoring Instrumentation and Controls .
* * * * -* . 3.2.2 Direct Exposure Determination  
2.2 Gaseous Effluent Monitor Setpoint Determination . .
****. i . 33
2.2.1 Containment and Plant Monitor 2.2.2 Conservative Default Values * * * * * . . .
* 34 . 35
2.3 Gaseous Effluent Instantaneous 15 17 17 19 Dose Rate Calculations - 10 CFR 20. * * * * *          . 20 2.3.1 Site Boundary Dose Rate - Noble Gases. .            20
* 35  
: 2. 3. 2 Site Boundary Dose Rate -                  .,
* *
Radioiodine and Particulates * * * * * :i . . 21 2.4 Noble Gas Effluent Dose Calculations - 10 c::FRt,50 . . 24 2.4.1 UNRESTRICTED AREA Dose - Noble Gases * * . .        24 2.4.2 Simplified Dose Calculation for Noble Gases.        25 2.5 Radioiodine and Particulate Dose Calculations -
* Salem ODCM Rev. 8 Table of Contents -continued
10 CFR 50. * * * * * * * * * * * * * * * * * . . .        26 2.5.1 UNRESTRICTED AREA Dose -
Radioiodine and Particulates * * * * * * . . 26 2.s.2 Simplified Dose Calculation for ~adioiodines and Particulates * * * * * * * * * * .        . 27
        -2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations * * * * * * * * * * * * . . .      28 2.7 Gaseous Effluent Dose Projection * * * * * * * . .        32 3.0  Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY. . 33 3.2 Doses to MEMBERS OF THE PUBLIC - 40 CFR 190 *
* 34
: 3. 2 .1 Effluent Dose Calculations * * * * -* .         . 35 3.2.2 Direct Exposure Determination * * * * .
* 35 i
* Table of Contents - continued 4.0 Salem ODCM Radiological Environmental Monitoring Program. .
4.1 Sampling Program . . * . * . . * * * * * * .
Rev. 8
                                                                          * .36
* 36 4.2 Interlaboratory Comparison Program * * . * .          *    *
* 37 Tables 1-1 Parameters for Liquid Alarm Setpoint Determination
              - Unit 1 .  . . . .  .  . . . . .  .  . .  . .  . .  .  .  . 41 1-2  Parameters for Liquid Alarm Setpoint Determination
              - Unit 2 . .  . . . . .  . . . . . .  . .  . .  .  . .    . 42 1-3  Site Related Ingestion Dose Commitment Factors, Am * * * * * * * * * * * * * * * ~ ... 43-44 1-4  Bioaccumulation Factors (BFi) * * * * * * * . . . . 45 2-1  Dose Factors for Noble Gases * * * * * * * * . . . 48 2-2  Parameters for Gaseous Alarm Setpoint Determinations 2-3
              - Unit 1 * * * * * * *    ...  * * * * * * * . * . . 49 Parameters for Gaseous Alarm Setpoint Determinations
              - Unit 2 . .  . . . . .  . . . ._ . .  . .  .. . . . .    . 50 2-4  Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations * * * * * * . . . 51 2-5 Pathway Dose Parameters - Atmospheric Releases 52-63
* A-1 A-2 B-1 B-2 Calculation of Effective MPC - Unit 1 * . . . . . A-5 Calculation of Effective MPC - Unit 2 * * * * *
* A-6 Adult Dose Contributions Fish Pathways Unit 1 * * *
* and Drinking water
                                          . . .and Adult Dose Contributions Fish
                                                . .Drinking Water B-5 Pathways Unit 2    *.* * * * *  * * * .* * * * * :_'r..
* B-5 C-5  Effective Dose Factors * * *      * * * * * * * * ,, * . C-6 Appendices Appendix A - Evaluation of Conservative, .Default MPC Value for Liquid Effluents * * * * . . A-1 Appendix B - Technical Basis for Effective Dose Factors -
Liquid Radioactive Effluents * * * . . B-1 Appendix c - Technical Bases for Effective Dose Factors -
Gaseous Radioactive Effluents              . . c-1
      **Appendix D - Radiological Environmental Monitoring Proqram - Sample Type, Location and Analysis * * * * * * * * * * * * * . D-1 ii


===4.0 Radiological===
-.                                            Salem ODCM SALEK NUCLEAR GENERATING STATION Rev. a OFFSITB DOSB CALCULATIOH MAHOAL Introduction The Salem Offsite Dose Calculation Manual (ODCM) describes the methodology and parameters used in: 1) the calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints; and 2) the calculation of radioactive liquid and gaseous concentrations, dose rates and cumulative quarterly and yearly doses. The methodology stated in this manual is acceptable for use in demonstrating compliance with 10 CFR 20.106, 10 CFR so, Appendix I and 40 CFR 190.
* More conservative calculation methods and/or conditions (e.g.,
location and/or exposure pathways) expected to yield higher computed doses than appropriate for the maximally exposed person may be assumed in the dose evaluations.
The ODCM will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations. Changes will be made to the ODCM calculation methodoloqies and parameters as is deemed necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR S0.36a and Appendix I for demonstrating  radi~active  effluents are ALARA.
NOTE.:  As used throughout this document, excluding acronyms, words appearing all capitalized denote the application of definitions as used in the Salem Technical Specifications .
* 1
* Salem ODCM  Rev. 8 1.0   Liquid Effluents 1.1 Radiation Monitoring Instrumentation and controls The liquid effluent monitoring instrumentation and .controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Salem Radiological Effluent Technical Specifications are summarized as follows: ,
: 1)  Alarm Cand Automatic Termination> - l-Rl8 (Unit 1) and 2-Ris (Unit 2) provide the alarm and automatic termination of liquid radioactive material releases as requir~d by *
* Technical Specification 3.3.3.8.
1-Rl9 A,B,C,and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R~9 A,B,C and D provide this function for Unit 2.
: 2)  Alarm Conly> - The alarm functions for the Service ~ter system are provided by the radiation monitors on the*
Containment Fan Cooler discharges (1-R 13 A,B,C,D and E for Unit 1 and 2-R 13 A,B,and c for Unit 2).
Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored (with an alarm function) by R37 prior to release to the environment.
Liquid radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented as Figures 1-1 and 1-2 for Units l and 2, respectively .
* 2


Environmental Monitoring Program. . 4.1 Sampling Program ..*.*..******.
Salem ODCM  Rev. 8 1.2  Liqui4 lffluent Monitor sotpoint Determination Per the requirements of Technical Specification 3.3.3.8, alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of Specification 3.11.1.1 are met (i.e., the concentration- of radioactive material released in liquid effluents to UNRESTRICTED AREAS.shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, for radionuclides and 2.oE-04* uCi/ml for dissolved or entrained noble gases). The f ollowinq equation* must be satisfied to meet the liquid effluent restrictions:
* where:
cs C  (F+f) f (1.1) c = the  effluent    concentration  limit    of    ~chnical Specification (3.11.1.1) implementing the 10 CFR 20 MPC for the site, in uCi/ml      .
c = the setpoint, in uCi/ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, represents a value which, if exceeded, would result in concentrations exceeding the limits of 10 CFR 20 in the UNRESTRICTED AREA f = the flow rate at the radiation monitor location, in volume per unit time, but in the same units as F, below F = the dilution water flow rate as measured prior to the release point, in volume per unit time
[Note that if no dilution is provided, cs c. Also, note that when (F) is large compared to (f), then (F + f) = F.]
* Adapted from NUREG-0133 3


===4.2 Interlaboratory===
Salem ODCM  Rev. a Liquid    Effluent  Monitors      CRadwaste, steam    Generator Blowdown. Chemical Waste Basin and service water.            The setpoints for the liquid effluent monitors at the Salem Nuclear Generating station are determined by the following equations:
MPCe
* SEN
* CW SP  S,                        + bkq        (1. 2)
RR with:
l:  Ci MPCe = ------                            (1. 3)
Ci
* where:
SP    =     alarm setpoint correspondinq to the maximum allowable release rate (cpm)
MPCe =       an effective MPC value for the mixture of *;:,
radionuclides in the effluent stream (uCi/mlL)
Ci    =    the concentration of radionuclide i in the undiluted liquid* effluents (uCi/ml)*
9NOTE:  The concentration mix must include the most recent composite of alpha emitters, sr-89, sr-90, Fe-55, and H-3 per Technical Specification J.11.1.1.
MP Ci =    the MPC value correspondinq to radionuclide i from 10 CFR 20, Appendix B, Table II, Column*2 (uCi/ml)
SEN      =   the sensitivity value to which the monitor is calibrated (cpm per uCi/ml)
* cw      =   the circulating water flow rate (dilution water flow) at the time of release (qal/min)
RR      =   the liquid effluent release rate (qal/min) bkg    =  the backqround of the monitor (cpm)
* 4
* Salem ODCM  Rev. a The radioactivity monitor setpoint equation (1.2) remains valid during outages when the circulating water dilution is potentially at its lowest value. Reduction of the waste stream flow (RR) may be necessary during these periods to meet the discharge criteria.
However, in order to maximize the available plant discharge dilution and thereby minimize the potential offsite doses, releases from either Unit-1 or Unit-2 may be routed to either the Unit-1 or Unit-2 Circulating Water System discharge. This routing is possible via interconnections between the Service Water Systems (see Figures 1 and 2). Procedural restrictions prevent simultaneous releases from either a single unit or both units into a single Circulating Water system
                                            . discharge.
* 1.2.2    conservative Default Values. Conserv~tive alarm setpoints may be determined through the use of default parameters.        Tables 1-1 and 1-2 summarize all current default values in use for Salem Unit-1 and Unit-2, respectively. They are based upon th,~.:
                                                              *-11 following:
a)  substitution of the effective MPC value with a default value of 4.71E-06 uci/ml (Unit 1) and 3.38E-06 uci/ml (Unit 2). (refer to Appendix A for justification);
b)  *or additional conservatism*, substit~tion of the I~131 MPC value of JE-07 uci/ml for the Ri9 steam Generator Slowdown Monitors, the R-37 Chemical waste Baain monitor and the R-13 Service Water Monitors.
* Based upon the potential for I-131 to be present in the secondary and service water systems, the use of the default effective MPC value as derived in Appendix A may be non-conservative for the 1, 2 R-19 SGBD monitors, the R-37 Chemical Waste Basin Monitor and the R-13 Service Water
* Monitors
* s


Comparison Program **.*. * * . 3 6 * *
                                                                      -~
* 3 6 * *
** c) d)
* 3 7 Tables 1-1 Parameters for Liquid Alarm Setpoint Determination
Salem ODCM  Rev. 8 substitutions of the operational circulating water flow with the lowest flow, in gal/min; and, substitutions of the effluent release rate with the highest allowed rate, in gal/min.
-Unit 1 . . . . . . . . . . . . . . . . . . . . .
With pre-established alarm setpoints, it is possible to control the radwaste release rate (RR) to ensure the inequality of equation {1.2) is maintained under changing values for MPGe and for differing Circulating Water System dilutions
41 1-2 Parameters for Liquid Alarm Setpoint Determination
* The Unit 2 Service Water system utilizes the Unit 1 Circulating Water-* systb for dilution prior to release to the river. It is possible to have the Unit 1 Circulating Water system. out of service when Unit 1 is in an outage. So, for conservatism no dilution is used for determining a 2R13 default alarm setpoint.
-Unit 2 . . . . . . . . . . . . . . . . . . . . .
Because no dilution is considered and the 2R13 monitor sensitivity is high, the MPCe of 3.JSE-06 uCi/ml is used in calculating the alarm setpoint (otherwise using 3E-07 uCi/ml would result in an alarm setpoint of 1 cpm) *
42 1-3 Site Related Ingestion Dose Commitment Factors, Am * * * * * * * * * * * * * * * ... 43-44 1-4 Bioaccumulation Factors (BFi) *******....
* 6
45 2-1 Dose Factors for Noble Gases * * * * * * * * . . . 48 2-2 Parameters for Gaseous Alarm Setpoint Determinations
* 1.3 Salem ODCM Liquit lffluent concentration Limits - 10 CPR 20 Rev. a Technical Specification 3.11.1.1 limits the concentration of radioactive material in liquid effluents (after dilution in the circulating Water System) to less than the concentrations as specified in 10 CFR 20, Appendix B, Table II, Column 2 for radionuclides other than noble gases. Noble gases are limited to a diluted concentration of 2.0E-04 uCi/ml. Release rates are controlled and radiation monitor alarm setpoints are established as addressed above to ensure that these concentration limits are not exceeded. However, in the event any liquid release results in an alarm setpoint being exceeded, an evaluation of compliance
-Unit 1 * * * * * *
** with the concentration limit's of Technical Specification 3.11.1.1 may be performed using the following equation:
* 2-3 2-4 2-5 A-1 A-2 B-1 B-2 C-5 Appendices . . . * * * * * * * . * . . 49 Parameters for Gaseous Alarm Setpoint Determinations
Ci          RR                ~-
-Unit 2 . . . . . . . . . .  
                                                              ~
._ . . . . .. . . . . .
                                                              \~
50 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations
E
* * * * * * . . . 51 Pathway Dose Parameters
                                        *             -~ 1       (1.4)
-Atmospheric Releases 52-63 Calculation of Effective MPC -Unit 1 * . . . . .
                              ~~          cw + RR where:
A-5 Calculation of Effective MPC -Unit 2 * * * * *
Ci  = actual  concentration of radionuclide i as measured in the undiluted liquid effluent (~Ci/ml)
* A-6 Adult Dose Contributions Pathways Unit 1 * * *
        ~~  =- the MPC value corresponding to radionuclide i from 10 CFR 20, Appendix B, Table II, Column 2 (uCi/ml)
* Fish and Drinking water . . . . . . . . . . B-5 Adult Dose Contributions Fish and Drinking Water Pathways Unit 2 * . * * * * * * * * .* * * * * :_'r..
              =- 2E-04 uCi/ml for dissolved or entrained noble gases RR    = the actual liquid effluent release rate (gal/min) cw    = the actual circulating water flow rate (dilution water flow) at the time of the release (gal/min)
* B-5 Effective Dose Factors * * * * * * * * * * * ,, * . C-6 Appendix A -Evaluation of Conservative, .Default MPC Value for Liquid Effluents
* 7
* * * * . . A-1 Appendix B -Technical Basis for Effective Dose Factors -Liquid Radioactive Effluents
* * * . . B-1 Appendix c -Technical Bases for Effective Dose Factors -Gaseous Radioactive Effluents . . c-1 **Appendix D -Radiological Environmental Monitoring Proqram -Sample Type, Location and Analysis * * * * * * * * * * * * * . D-1 ii 
-. *
* Salem ODCM Rev. a SALEK NUCLEAR GENERATING STATION OFFSITB DOSB CALCULATIOH MAHOAL Introduction The Salem Offsite Dose Calculation Manual (ODCM) describes the methodology and parameters used in: 1) the calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints; and 2) the calculation of radioactive liquid and gaseous concentrations, dose rates and cumulative quarterly and yearly doses. The methodology stated in this manual is acceptable for use in demonstrating compliance with 10 CFR 20.106, 10 CFR so, Appendix I and 40 CFR 190. More conservative calculation methods and/or conditions (e.g., location and/or exposure pathways) expected to yield higher computed doses than appropriate for the maximally exposed person may be assumed in the dose evaluations.
The ODCM will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations.
Changes will be made to the ODCM calculation methodoloqies and parameters as is deemed necessary to ensure . ' reasonable conservatism in keeping with the principles of 10 CFR S0.36a and Appendix I for demonstrating effluents are ALARA. NOTE.: As used throughout this document, excluding acronyms, words appearing all capitalized denote the application of definitions as used in the Salem Technical Specifications . 1 
* *
* Salem ODCM Rev. 8 1.0 Liquid Effluents


===1.1 Radiation===
Salem ODCM    Rev. a
~* 1.4  Liqui4 lffluent Dose Calculation - 10 Cl'R so 1.4.1  MEMBER OP THE PUBLIC Dose - Liquid Effluents.
Technical  Specific~tion  3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBLIC from radioactive materials in liquid effluents from each unit of the Salem Nuclear Generating Station to:
                      - during any calendar quarter;
                    ~  105 mrem to total body per unit
                    ~  5.0 mrem to any organ per unit
                    - during any calendar year;
                    ~  3.0 mrem to total body per unit
                    ~  10.0 mrem to any organ per unit.
Per the surveillance requirements of Technical Specification 4.11.1.2, the following calculation methods shall be used for
* determining the dose or dose commitment due to the liquid radioactive effluents from Salem.
l.67E-02
* VOL
( 1. 9'l cw where:
dose or dose commitment to organ o (mrem). Total body dose can also be calculated using site- related total body dose commitment factor.
site-related ingestion dose commitment factor to the total body or any organ o for radionuclide i (mrem/hr per uCi/ml)                        -
* average concentration of radionuclide i, in undiluted liquid effluent representative of the volume VOL *
(uCi/ml)
VOL=    volume of liquid effluent released (gal) cw =    average circulating water discharge rate during release period (gal/min)
: 1. 67E-02 =  conversion factor (hr/min)
**                                    8
* Salem ODCM  Rev. 8 The site-related ingestion dose/dose commitment factors (Aw) are presented in Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:
( 1. 6) where:
Aw =    composite dose parameter for the total body or critical organ o of an adult for radionuclide i, for the fish and invertebrate ingestion pathways (mrem/hr per uci/ml)
UI =    adult invertebrate consumption (5 kg/yr)
Bii =  bioaccumulation factor for radionuclide*i in invertebrates from Table 1-4 (pCi/kg per pCi/l)
UF =   adult fish consumption (21 kg/yr)
BFi =   bioaccumulation factor for radionuclide i in fish from Table 1-4 (pCi/kg per pCi/l) dose conversion factor for nuclide i for adults in pre~
selected organ, o, from Table E-11 of Regulatory Guide 1.109 (mrem/pCi)                                    -
1.14E+05=    conversion factor (pCi/uCi
* ml/kg per hr/yr)
The radionuclides included in the periodic dose  assessmen~  per the requirements of Technical Specification 3/4.11.1.2      are.''tho~e as identified by gamma spectral analysis of the liquid waste samples collected  and  analyzed  per  the  requirements  of  Technical Specification 3/4.11.1.1, Table 4.11-1.
Radionuclid- requiring radiochemical analysis (e.g., Sr-89 and sr-
: 90) will be added to the dose analysts at a frequency consistent with  the required    minimum  analysis  frequency  of  Technical Specification Table 4.11-1 .
**                                  9
* Salem ODCM 1.4.2 Simplified Liquid Effluent Dose Calculation.
Rev. a In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following simplified dose calculation equation may be used for demonstrating compliance with the dose limits of Technical Specification 3.11.1.2.        (Refer to Appendix B for the derivation and justification for this simplified method.)
Total Body 1.21E+03
* VOL D111 =                               (1.7) cw
~ Maximum Organ 2.52E+04
* VOL Dmu  =                            (1. 8) cw where:
q    =     average concentration of radionuclide i, in undiluted liquid effluent representative of the volume VOL (uCi/ml)
VOL  =     volwaa of liquid effluent released (gal) cw  -~
average circulating water dischar9e r~te during release period (gal/min)
* Dlb o_  ----- conservatively evaluated total body dose (mrem) conservatively evaluated maximum organ dose (mrem)
: 1. 21E+03  =     conversion factor (hr/min) and the conservative total body dose conversion factor (Fe-59, total body -- 7.27E+04 mrem/hr per uCi/ml) 2.52E+04  =    conversion factor (hr/min) and the conservative maximum organ dose conversion factor (Nb-95, GI-LLI
                  -- 1.51E+06 mrem/hr per uCi/ml)
* 10
* Salem ODCM 1.s secondary Sid* Radioactive Liquid Effluents and Dose Calculat~ODI  Qurinq Primary    to Secondary Leakage Rev. 8 During  periods  of primary to    secondary  leakage  (i.e.,      steam generator tube leaks), radioactive material will be transmitted from the primary system to the secondary system.        The potential exists for the release of radioactive material to the off-site environment  (Delaware' River)  via secondary system dischargeso Potentially significant radioactive material levels and potential releases are controlled/monitored by the    Stea~  Generator blowdown monitors (Rl9) and the Chemical Waste Basin monitor (R37).        However to ensure compliance with the regulatory limits on radioactive material releases, it may be desirable to account for potential releases from the secondary system during periods of primary to secondary leakage. Any potentially significant releases will be via the Chemical Waste Basin with the major source        of*~activity
                                                              . '((.
being the Steam Generator blowdown.
With  identified radioactive material      levels  in the    secondary system, appropriate samples should be collected and analyzed for the  principal  gamma  emitting  radionuclides~    Based      on    the identified radioactive material levels and the volume of water discharged, the resulting environmental doses may be calculated based on equation (1.5) .
**                                    11


Monitoring Instrumentation and controls The liquid effluent monitoring instrumentation and .controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Salem Radiological Effluent Technical Specifications are summarized as follows: , 1) Alarm Cand Automatic Termination>
Salem ODCM   Rev. a Because the release rate from the secondary system is indirect (e.g. , SG blowdown is normally routed to condenser where the condensate clean-up system will remove much of the radioactive material), samples should be collected from the final release point (i.e.,   Chemical Waste Basin) for   quantifying the radioactive material   releases. However,   for conservatism   and   ease of controlling and quantifying all potential release paths,       it is prudent to sample the SG blowdown and to assume all radioactive material is released directly to the environment via the Chemical waste Basin. This approach while not exact, is conservative and ensures timely analysis for regulatory compliance. Accounting for radioactive material retention of the condensate clean-up system
-l-Rl8 (Unit 1) and 2-Ris (Unit 2) provide the alarm and automatic termination of liquid radioactive material releases as by
* ion exchange resins may be needed to more accurately account for actual releases *
* Technical Specification 3.3.3.8. 1-Rl9 A,B,C,and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines.
* 12
A,B,C and D provide this function for Unit 2. 2) Alarm Conly> -The alarm functions for the Service system are provided by the radiation monitors on the* Containment Fan Cooler discharges (1-R 13 A,B,C,D and E for Unit 1 and 2-R 13 A,B,and c for Unit 2). Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored (with an alarm function) by R37 prior to release to the environment.
Liquid radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented as Figures 1-1 and 1-2 for Units l and 2, respectively . 2
* Salem ODCM Rev. 8 1.2 Liqui4 lffluent Monitor sotpoint Determination Per the requirements of Technical Specification 3.3.3.8, alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of Specification 3.11.1.1 are met (i.e., the concentration-of radioactive material released in liquid effluents to UNRESTRICTED AREAS.shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, for radionuclides and 2.oE-04* uCi/ml for dissolved or entrained noble gases). The f ollowinq equation*
must be satisfied to meet the liquid effluent restrictions:
where: c = c = C (F+f) (1.1) cs f the effluent concentration limit of Specification (3.11.1.1) implementing the 10 CFR 20 MPC for the site, in uCi/ml . the setpoint, in uCi/ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, represents a value which, if exceeded, would result in concentrations exceeding the limits of 10 CFR 20 in the UNRESTRICTED AREA f = the flow rate at the radiation monitor location, in volume per unit time, but in the same units as F, below F = the dilution water flow rate as measured prior to the release point, in volume per unit time [Note that if no dilution is provided, cs c. Also, note that when (F) is large compared to (f), then (F + f) = F.]
* Adapted from NUREG-0133 3 
*
* Salem ODCM Rev. a Liquid Effluent Monitors CRadwaste, steam Generator Blowdown.
Chemical Waste Basin and service water. The setpoints for the liquid effluent monitors at the Salem Nuclear Generating station are determined by the following equations:
with: where: SP = MPCe = Ci = 9NOTE: MP Ci = SEN = cw = RR = bkg = MPCe
* SEN
* CW SP S, + bkq (1. 2) RR l: Ci MPCe = ------(1. 3) Ci alarm setpoint correspondinq to the maximum allowable release rate (cpm) an effective MPC value for the mixture of *;:, radionuclides in the effluent stream (uCi/mlL) the concentration of radionuclide i in the undiluted liquid* effluents (uCi/ml)*
The concentration mix must include the most recent composite of alpha emitters, sr-89, sr-90, Fe-55, and H-3 per Technical Specification J.11.1.1.
the MPC value correspondinq to radionuclide i from 10 CFR 20, Appendix B, Table II, Column*2 (uCi/ml) the sensitivity value to which the monitor is calibrated (cpm per uCi/ml)
* the circulating water flow rate (dilution water flow) at the time of release (qal/min) the liquid effluent release rate (qal/min) the backqround of the monitor (cpm) 4
* Salem ODCM Rev. a The radioactivity monitor setpoint equation (1.2) remains valid during outages when the circulating water dilution is potentially at its lowest value. Reduction of the waste stream flow (RR) may be necessary during these periods to meet the discharge criteria.
However, in order to maximize the available plant discharge dilution and thereby minimize the potential offsite doses, releases from either Unit-1 or Unit-2 may be routed to either the Unit-1 or Unit-2 Circulating Water System discharge.
This routing is possible via interconnections between the Service Water Systems (see Figures 1 and 2). Procedural restrictions prevent simultaneous releases from either a single unit or both units into a single Circulating Water system discharge. . .
* 1.2.2 conservative Default Values.
alarm setpoints may be determined through the use of default parameters.
Tables 1-1 and 1-2 summarize all current default values in use for Salem
* Unit-1 and Unit-2, respectively.
following:
They are based upon
*-11 a) b) substitution of the effective MPC value with a default value of 4.71E-06 uci/ml (Unit 1) and 3.38E-06 uci/ml (Unit 2). (refer to Appendix A for justification);
*or additional conservatism*,
of the MPC value of JE-07 uci/ml for the Ri9 steam Generator Slowdown Monitors, the R-37 Chemical waste Baain monitor and the R-13 Service Water Monitors.
* Based upon the potential for I-131 to be present in the secondary and service water systems, the use of the default effective MPC value as derived in Appendix A may be conservative for the 1, 2 R-19 SGBD monitors, the R-37 Chemical Waste Basin Monitor and the R-13 Service Water Monitors
* s 
** *
* Salem ODCM Rev. 8 c) substitutions of the operational circulating water flow with the lowest flow, in gal/min; and, d) substitutions of the effluent release rate with the highest allowed rate, in gal/min. With pre-established alarm setpoints, it is possible to control the radwaste release rate (RR) to ensure the inequality of equation {1.2) is maintained under changing values for MPGe and for differing Circulating Water System dilutions
* The Unit 2 Service Water system utilizes the Unit 1 Circulating Water-* systb for dilution prior to release to the river. It is possible to have the Unit 1 Circulating Water system. out of service when Unit 1 is in an outage. So, for conservatism no dilution is used for determining a 2R13 default alarm setpoint.
Because no dilution is considered and the 2R13 monitor sensitivity is high, the MPCe of 3.JSE-06 uCi/ml is used in calculating the alarm setpoint (otherwise using 3E-07 uCi/ml would result in an alarm setpoint of 1 cpm)
* 6 
* **
* Salem ODCM Rev. a 1.3 Liquit lffluent concentration Limits -10 CPR 20 Technical Specification 3.11.1.1 limits the concentration of radioactive material in liquid effluents (after dilution in the circulating Water System) to less than the concentrations as specified in 10 CFR 20, Appendix B, Table II, Column 2 for radionuclides other than noble gases. Noble gases are limited to a diluted concentration of 2.0E-04 uCi/ml. Release rates are controlled and radiation monitor alarm setpoints are established as addressed above to ensure that these concentration limits are not exceeded.
However, in the event any liquid release results in an alarm setpoint being exceeded, an evaluation of compliance with the concentration limit's of Technical Specification 3.11.1.1 may be performed using the following equation:
where: Ci = =-=-RR = cw = Ci RR E
* 1 ;; (1.4) cw + RR actual concentration of radionuclide i as measured in the undiluted liquid effluent the MPC value corresponding to radionuclide i from 10 CFR 20, Appendix B, Table II, Column 2 (uCi/ml) 2E-04 uCi/ml for dissolved or entrained noble gases the actual liquid effluent release rate (gal/min) the actual circulating water flow rate (dilution water flow) at the time of the release (gal/min) 7 Salem ODCM Rev. a 1.4 Liqui4 lffluent Dose Calculation
-10 Cl'R so 1.4.1 MEMBER OP THE PUBLIC Dose -Liquid Effluents.
Technical 3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBLIC from radioactive materials in liquid effluents from each unit of the Salem Nuclear Generating Station to: -during any calendar quarter; 105 mrem to total body per unit 5.0 mrem to any organ per unit -during any calendar year; 3.0 mrem to total body per unit 10.0 mrem to any organ per unit. Per the surveillance requirements of Technical Specification 4.11.1.2, the following calculation methods shall be used for
* determining the dose or dose commitment due to the liquid radioactive effluents from Salem. ** where: VOL= cw = 1. 67E-02 = l.67E-02
* VOL ( 1. 9'l :.;!* cw dose or dose commitment to organ o (mrem). Total body dose can also be calculated using site-related total body dose commitment factor. site-related ingestion dose commitment factor to the total body or any organ o for radionuclide i (mrem/hr per uCi/ml) -** average concentration of radionuclide i, in undiluted liquid effluent representative of the volume VOL * (uCi/ml) volume of liquid effluent released (gal) average circulating water discharge rate during release period (gal/min) conversion factor (hr/min) 8 
* * ** Salem ODCM Rev. 8 The site-related ingestion dose/dose commitment factors (Aw) are presented in Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:
where: Aw = UI = Bii = UF = BFi = 1.14E+05= ( 1. 6) composite dose parameter for the total body or critical organ o of an adult for radionuclide i, for the fish and invertebrate ingestion pathways (mrem/hr per uci/ml) adult invertebrate consumption (5 kg/yr) bioaccumulation factor for radionuclide*i in invertebrates from Table 1-4 (pCi/kg per pCi/l) adult fish consumption (21 kg/yr) bioaccumulation factor for radionuclide i in fish from Table 1-4 (pCi/kg per pCi/l) dose conversion factor for nuclide i for adults in selected organ, o, from Table E-11 of Regulatory Guide 1.109 (mrem/pCi)
-conversion factor (pCi/uCi
* ml/kg per hr/yr) The radionuclides included in the periodic dose per the requirements of Technical Specification 3/4.11.1.2 as identified by gamma spectral analysis of the liquid waste samples collected and analyzed per the requirements of Technical Specification 3/4.11.1.1, Table 4.11-1. Radionuclid-requiring radiochemical analysis (e.g., Sr-89 and sr-90) will be added to the dose analysts at a frequency consistent with the required minimum analysis frequency of Technical Specification Table 4.11-1 . 9
* Salem ODCM Rev. a 1.4.2 Simplified Liquid Effluent Dose Calculation.
In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following simplified dose calculation equation may be used for demonstrating compliance with the dose limits of Technical Specification 3.11.1.2. (Refer to Appendix B for the derivation and justification for this simplified method.) Total Body 1.21E+03
* VOL D111 = (1.7) cw Maximum Organ where: q = VOL = cw Dlb --o_ ---1. 21E+03 = 2.52E+04 =
* 2.52E+04
* VOL Dmu = (1. 8) cw average concentration of radionuclide i, in undiluted liquid effluent representative of the volume VOL (uCi/ml) volwaa of liquid effluent released (gal) average circulating water dischar9e during release period (gal/min)
* conservatively evaluated total body dose (mrem) conservatively evaluated maximum organ dose (mrem) conversion factor (hr/min) and the conservative total body dose conversion factor (Fe-59, total body --7.27E+04 mrem/hr per uCi/ml) conversion factor (hr/min) and the conservative maximum organ dose conversion factor (Nb-95, GI-LLI --1.51E+06 mrem/hr per uCi/ml) 10 
* * ** Salem ODCM Rev. 8 1.s secondary Sid* Radioactive Liquid Effluents and Dose Qurinq Primary to Secondary Leakage During periods of primary to secondary leakage (i.e., steam generator tube leaks), radioactive material will be transmitted from the primary system to the secondary system. The potential exists for the release of radioactive material to the off-site environment (Delaware' River) via secondary system dischargeso Potentially significant radioactive material levels and potential releases are controlled/monitored by the Generator blowdown monitors (Rl9) and the Chemical Waste Basin monitor (R37). However to ensure compliance with the regulatory limits on radioactive material releases, it may be desirable to account for potential releases from the secondary system during periods of primary to secondary leakage. Any potentially significant releases will be via the Chemical Waste Basin with the major source . '{{. being the Steam Generator blowdown.
With identified radioactive material levels in the secondary system, appropriate samples should be collected and analyzed for the principal gamma emitting Based on the identified radioactive material levels and the volume of water discharged, the resulting environmental doses may be calculated based on equation (1.5) . 11 
*
* Salem ODCM Rev. a Because the release rate from the secondary system is indirect (e.g. , SG blowdown is normally routed to condenser where the condensate clean-up system will remove much of the radioactive material), samples should be collected from the final release point (i.e., Chemical Waste Basin) for quantifying the radioactive material releases.
However, for conservatism and ease of controlling and quantifying all potential release paths, it is prudent to sample the SG blowdown and to assume all radioactive material is released directly to the environment via the Chemical waste Basin. This approach while not exact, is conservative and ensures timely analysis for regulatory compliance.
Accounting for radioactive material retention of the condensate clean-up system .. ion exchange resins may be needed to more accurately account for actual releases
* 12
* ** Salem ODCM Rev. a 1.6 Liqui4 lffluent Dose Pro1ections Technical Specification 3.11.1.3 requires that the liquid radioactive waste processing system be used to reduce the radioactive material levels in the liquid waste prior to release when the quarterly projected doses exceed: 0.375 mrem to the total body, or 1.25 mrem to any organ. The* applicable liquid waste processing system for maintaining radioactive material releases ALARA is the ion exchange system as delineated in Figure 1-3. Alternately, the waste evaporator as presented in the Salem FSAR.has processing capabilities meeting the NRC ALARA design requirements and may be used in conjunction or in lieu of the ion exchange system for waste processing requirements in accordance with Technical Specification 3.11.1.3.
These processing requirements are applicable to each unit indiV'tidually.
*i. Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing requirements for the other unit
* 13 
-* *
* Salem ODCM Rev. 8 Dose projections are made at least once per 31 days by the following equations:
where: Dtbp = Dtb = om.up = Dmu = d = 91 = Dtbp = Dmaxp = Dlb (91 I d) Dmu (91 I d) (1. 9) ( 1. 10) the total body dose projection for current calendar quarter (mrem) *
* the total body dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) the maximum organ dose projection for current calendar quarter (mrem) the maximum organ dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) the number of days to date for current calendar quarter the number of days in a calendar quarter 14 
* * ** Salem ODCM Rev. a 2.0 Gas90u1 Bffluents


===2.1 Radiation===
Salem ODCM  Rev. a 1.6  Liqui4 lffluent Dose Pro1ections Technical  Specification  3.11.1.3  requires  that  the    liquid radioactive  waste  processing  system  be  used  to  reduce    the radioactive material levels in the liquid waste prior to release when the quarterly projected doses exceed:
0.375 mrem to the total body, or 1.25 mrem to any organ.
The* applicable liquid waste processing system for maintaining radioactive material releases ALARA is the ion exchange system as delineated in Figure 1-3. Alternately, the waste evaporator as presented in the Salem FSAR.has processing capabilities meeting the
* NRC ALARA design requirements and may be used in conjunction or in lieu of the ion exchange system for waste processing requirements in  accordance  with  Technical  Specification  3.11.1.3.      These processing requirements are applicable to each unit indiV'tidually.
                                                              *i.
Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing requirements for the other unit *
**                                  13


Monitoring Instrumentation and Controls The gaseous effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Radiological Effluent Technical Specifications are summarized as follows: 1) Waste Gas Holdup System -The vent header gases are collected by the waste gas holdup system. Gases may be recycled to provide cover gas for the eves hold-up tank or held in the waste gas tanks for decay prior to release. waste gas decay tanks are batch *released after sampling and analysis.
-*                                                Salem ODCM Dose projections are made at least once per 31 days following equations:
The tanks are discharged via the Plant Vent. l-R41C provides noble gas monitoring and automatic isolation of waste gas decay tank releases for Unit-1. This function is provided by 2-R41C for Unit-2. 2) Containment Purge and Pressure/Vacuum Relief -containment purges and pressure/vacuum reliefs are released to the atmosphere via the respective unit Plant Vent. Noble gas monitoring and auto isolation function are provided by l-R41C for Unit-1 and 2-R41C for Unit-2. Additionally, in accordance with Technical Specification
Rev. 8 by the Dtbp  =  Dlb (91 I d)              (1. 9)
: 3. 3. 3. 9, Table 3. 3-13, 1-Rl2A and 2-Rl2A may be used to provide the containment monitoring and automatic isolation function during purge. and reliefs.*  
Dmaxp  =  Dmu (91  I d)            ( 1. 10) where:
* :\: ?1 3) Plant Vent -The Plant Vent for each respective unit receives discharges from the waste gas hold-up system, condenser evacuation system, containment purge and pressure/vacuum reliefs, and the Auxiliary Building ventilation.
Dtbp  = the total body dose projection for current calendar quarter (mrem)                                    *
Effluents are monitored by R41C, a flow through gross activity monitor (for noble gas monitoring).
* Dtb  = the total body dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) om.up = the maximum organ dose projection for current calendar quarter (mrem)
Dmu  = the maximum organ dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) d    = the number of days to date for current calendar quarter 91    = the number of days in a calendar quarter
* 14
* 2.0 2.1 Gas90u1 Bffluents Salem ODCM Radiation Monitoring Instrumentation and Controls Rev. a The gaseous effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Radiological Effluent Technical Specifications are summarized as follows:
: 1) Waste Gas Holdup System - The vent header gases are collected by the waste gas holdup system.           Gases may be recycled to provide cover gas for the eves hold-up tank or held in the waste gas tanks for decay prior to release. waste gas decay tanks are batch *released after sampling and analysis. The tanks are discharged via the Plant Vent.
l-R41C provides noble gas monitoring and automatic isolation of waste gas decay tank releases for Unit-1. This function is provided by 2-R41C for Unit-2.
: 2) Containment Purge and Pressure/Vacuum Relief - containment purges and pressure/vacuum reliefs are released to the atmosphere via the respective unit Plant Vent.           Noble gas monitoring and auto isolation function are provided by l-R41C for Unit-1 and 2-R41C for Unit-2. Additionally, in accordance with Technical Specification 3. 3. 3. 9, Table 3. 3-13, 1-Rl2A and 2-Rl2A may be used to provide the containment monitoring and automatic isolation function during purge. and pressu~e/vacuum reliefs.*     *                                           :\:
                                                                ?1
: 3) Plant Vent - The Plant Vent for each respective unit receives discharges from the waste gas hold-up system, condenser   evacuation   system,     containment     purge     and pressure/vacuum   reliefs,     and   the   Auxiliary       Building ventilation. Effluents are monitored by R41C, a flow through gross   activity   monitor   (for   noble   gas   monitoring).
Additionally, in-line gross activity monitors, (1-R16 and
Additionally, in-line gross activity monitors, (1-R16 and
* The R12A monitors also provide the safety function of containment isolation in the event of a fuel handling accident during refueling.
* The R12A monitors also provide the safety function of containment isolation in the event of a fuel handling accident during refueling. During MODE 6 in accordance with Technical Specification 3/4.3.3, Table 3.3-6, the R12A alarm/trip setpoint shall be established at twice background, providing early indication and containment isolation accompanying unexpected increases in containment airborne radioactive material levels indicative of a fuel degradation. The R41C monitor may also provide
During MODE 6 in accordance with Technical Specification 3/4.3.3, Table 3.3-6, the R12A alarm/trip setpoint shall be established at twice background, providing early indication and containment isolation accompanying unexpected increases in containment airborne radioactive material levels indicative of a fuel degradation.
** this function if the R12A monitor is inoperable during MODE 6 .
The R41C monitor may also provide this function if the R12A monitor is inoperable during MODE 6 . 15
15
* *
* 3)
* Salem ODCM Rev. 8 3) Plant Vent Ccont'dl Rl6) provide redundant back-up monitoring capabilities to the R41C monitors.
Salem ODCM Rev. 8 Plant Vent Ccont'dl Rl6) provide redundant back-up monitoring capabilities to the R41C monitors. Radioiodine and particulate sampling capabilities are provided by charcoal cartridge and filter medium samplers with redundant back-up sampling capabilities provided by R41B and R41A, respectively. Plant Vent flow rate is measured and as a back-up may be determined empirically as a function of fan operation (fan curves). Sampler flow rates are determined by flow rate instrumentation (e.g., venturi rotameter) .
Radioiodine and particulate sampling capabilities are provided by charcoal cartridge and filter medium samplers with redundant back-up sampling capabilities provided by R41B and R41A, respectively.
Gaseous radioactive waste   flow   diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented as Figures 2-1 and 2-2 for Units 1 and 2, respectively .
Plant Vent flow rate is measured and as a back-up may be determined empirically as a function of fan operation (fan curves). Sampler flow rates are determined by flow rate instrumentation (e.g., venturi rotameter) . Gaseous radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented as Figures 2-1 and 2-2 for Units 1 and 2, respectively . 16
* 16
* Salem ODCM Rev. a 2.2 Gaseoq1 Effluent Monitor setpoint Determination
* 2.2
: 2. 2. 1 containment and Plant vent Monitor. Per the requirements of Technical Specification 3.3.3.9, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of Specification 3.11.2.1, which corresponds to a dose rate at the SITE BOUNDARY of 500 mrem/year to the total body or 3000 mrem/year to the skin. Based on a grab sample analysis of the applicable release (i.e., grab sample of the Containment atmosphere, waste gas decay tank, or Plant Vent), the radiation monitoring alarm setpoints may be established by the following calculation method. The measured radionuclide concentrations and ..
: 2. 2. 1 Gaseoq1 Effluent Monitor setpoint Determination containment and Plant vent Monitor.
* release rate are used to calculate the of the allowable release rate, as limited by Specification J.11.2.1, by the
Salem ODCM    Rev.
* equation:
Per the requirements of a
Technical     Specification   3.3.3.9,       alarm   setpoints     shall     be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of Specification 3.11.2.1, which corresponds to a dose rate at the SITE BOUNDARY of 500 mrem/year to the total body or 3000 mrem/year to the skin.         Based on a grab sample analysis of the applicable     release   (i.e.,   grab     sample   of   the   Containment atmosphere, waste gas decay tank, or Plant Vent), the radiation monitoring alarm setpoints may be established by the following calculation method.       The measured
                                    .        radionuclide concentrations and
* release rate are used to calculate the frac~ion of the allowable release     rate,   as limited     by   Specification     J.11.2.1,   by   the equation:
FRAC = [ 4. 72E+02
FRAC = [ 4. 72E+02
* X/Q
* X/Q
* VF
* VF
* l: (Ci * ] / 500 ( 2. 1) FRAC = [4.72E+02
* l: (Ci * ~) ] / 500                 ( 2. 1)
* X/Q *VF* I:
FRAC = [4.72E+02
1.1  
* X/Q *VF* I:           {Ci*{~+    1.1   ~))]  / 3000   (2.2) where FRAC = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate                                           .
/ 3000 (2.2) where FRAC X/Q VF = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate . = annual average meteorological dispersion to the controlling site boundary location {sec/m 3) = ventilation system flow rate for the applicable release point and monitor (ft 3/min) = concentration of noble gas radionuclide i as determine radioanalysis of grab sample {uCi/cm3)  
X/Q    = annual average meteorological dispersion 3to the controlling site boundary location {sec/m )
= total body dose conversion factor for noble gas radionuclide i (mrem/yr per uci/m 3 from Table 2-1) = beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) 17   
VF      = ventilation system flow       rate for the applicable release point and monitor (ft3 /min)
* * ** Salem ODCM Rev. a = gamma air dose conversion factor for noble gas 1.1 500 3000 4.72 radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) = mrem skin dose per mrad gamma air dose (mrem/mrad)  
          = concentration of noble gas radionuclide i as determine radioanalysis of grab sample {uCi/cm3)
= total body dose rate limit (mrem/yr)  
          = total body dose conversion factor 3 for noble gas radionuclide i (mrem/yr per uci/m from Table 2-1)
= skin dose rate limit (mrem/yr)
          = beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m3 from Table 2-1)
E+02 = conversion factor (cm /ft 3
* 17
* min/sec) Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoints for the applicable monitors (R16, R41C, and/or R12A) may be calculated by the equation:
 
where: SP SEN bkg AF = = = = SP = [AF
  *        =
Salem ODCM gamma air dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m3 from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad)
Rev. a 500 = total body dose rate limit (mrem/yr) 3000 = skin dose rate limit (mrem/yr) 4.72 E+02 = conversion factor (cm /ft3
* min/sec)
Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoints for the applicable monitors (R16, R41C, and/or R12A) may be calculated by the equation:
SP = [AF
* I: Ci
* I: Ci
* SEN / FRAC] + bkg alarm setpoint corresponding to the maximum allowable release rate (cpm) monitor sensitivity (cpm per uCi/cm 3) background of the monitor (cpm) administrative allocation factor for the monitor and type release, which corresponds fraction of the total allowable release rate administratively allocated to the release. ( 2. 3) specific to the that is The allocation factor (AF) is an administrative control to ,. .. \; ensure that combined releases from Salem Uni ts 1 and 2 '''and Hope Creek will not exceed the regulatory limits on release rate from the site {i.e., the release rate limits of Technical Specification 3.11.2.1).
* SEN / FRAC] + bkg             ( 2. 3) where:
Normally, the combined AF value for Salem Units 1 and 2 is equal to o
SP  =  alarm setpoint corresponding to the maximum allowable release rate (cpm)
* 5 ( o. 25 per unit) , with the
SEN =  monitor sensitivity (cpm per uCi/cm3 )
* remainder
bkg =  background of the monitor (cpm)
: o. 5 allocated to Hope Creek. Any increase in AF above 0.5 for the Salem Nuclear Generating station will be coordinated with the Hope Creek Generating Station to ensure that the combined allocation factors for all units do not exceed 1.0
AF =  administrative allocation factor for the specific monitor and type release, which corresponds to the fraction of the total allowable release rate that is administratively allocated to the release.
* 18
The allocation factor (AF) is an administrative control     i~posed        to
* Salem ODCM Rev. a 2.2.2 Cog1eryative Default Values. A conservative alarm setpoint can be established, in lieu of the individual radionuclide evaluation based on the grab sample analysis, to eliminate the potential of periodically having to adjust the setpoint to reflect minor changes in radionuclide distribution and variations in release flow rate. The alarm setpoint may be conservatively determined by the default values presented in Table 2-2 and 2-3 for Units 1 and 2, respectively.
                                                                ,...\;
These values are based upon: the maximum ventilation (or purge) flow rate; a radionuclide distribution*
ensure that combined releases from Salem Uni ts 1 and 2 '''and Hope Creek will not exceed the regulatory limits on release rate from the site {i.e., the release rate limits of Technical Specification 3.11.2.1). Normally, the combined AF value for Salem Units 1 and 2 is equal to o
comprised of 95% Xe-133, 2% Xe-135, 1% Xe-133m, 1% Kr-88 and 1% Kr-85; and -an administrative allocation factor of 0.25 to.conservatively ensure that any simultaneous releases from Salem Units 1 and 2 do not exceed the maximum allowable release rate.
* 5     ( o. 25 per unit) , with the
* For this radionuclide distribution, the alarm setpoint based oh the total body dose rate is more restrictive than the corresponding setpoint based on the skin dose rate. The resulting conservative,  
* remainder o. 5 allocated to Hope Creek.       Any increase in AF above 0.5 for the Salem Nuclear Generating station will be coordinated with the Hope Creek Generating Station to ensure that the combined allocation factors for all units do not exceed 1.0 *
* .'{\-,: default setpoints are presented in Tables 2-2 and 2-3. :.;. Adopted from ANSI N237-1976/ANS-18.1, Source Term Specif_ications, Table 6 19 -*.::'.; 
**                                      18
* *
* 2.2.2 can  be Cog1eryative Default Values.
* Salem ODCM Rev. 8 2.3 GastOg* lffluent Instantaneous Dose Rate Calculations  
established, in lieu of the Salem ODCM individual Rev. a A conservative alarm setpoint radionuclide evaluation based on the grab sample analysis, to eliminate the potential of periodically having to adjust the setpoint to reflect minor   changes   in radionuclide distribution and     variations   in release flow rate.     The alarm setpoint may be conservatively determined by the default values presented in Table 2-2 and 2-3 for Units 1 and 2, respectively. These values are based upon:
-10 CD 20 2.3.1 sit* Boundary Dose Rate Noble Gases. Technical Specification 3.11.2.la limits the dose rate at the SITE BOUNDARY due to noble gas releases to mrem/yr, total body and mrem/yr, skin. Radiation monitor alarm setpoints are established to ensure that these release limits are not exceeded.
the maximum ventilation (or purge) flow rate; a radionuclide distribution* comprised of 95% Xe-133, 2%
In the event any gaseous releases from the station results in an alarm setpoint being exceeded, an evaluation of the SITE BOUNDARY dose rate resulting from the release shall be performed using the following equations:
Xe-135, 1% Xe-133m, 1% Kr-88 and 1% Kr-85; and
where: 1.1 ( 2. 4) and . D 1 = X/Q
    -   an administrative allocation factor of 0.25 to.conservatively ensure that any simultaneous releases from Salem Units 1 and 2 do not exceed the maximum allowable release rate.
* l: ( +
* For this radionuclide distribution, the alarm setpoint based oh the total body dose rate is more restrictive than the corresponding setpoint based on the skin dose rate. The resulting conservative,
* Qi) (2. 5) = total body dose rate (mrem/yr)  
                                                                ..:~
., = skin dose rate (mrem/yr)
                                                              .'{\-,:
_} = atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m 3) = average release rate of radionuclide i over the release_ period under evaluation (uCi/sec)  
default setpoints are presented in Tables 2-2 and 2-3.     :.;.
= total body dose conversion factor for noble gas radionuclide i (mrem/yr per uci/m 3 , from Table 2-1) = beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 ,
Adopted from ANSI N237-1976/ANS-18.1, Source Term Specif_ications, Table 6
Table 2-1) -q&J11Da air dose conversion factor for noble gas radionuclide i (mrad/yr per uci/m 3 , from Table 2-1) = mrem skin dose per mrad gamma air dose (mrem/mrad)
* 19
As appropriate, simultaneous releases from sa_lem Units 1 and 2 and Hope creek will be considered in evaluating compliance with the releas*e rate limits of Specification 3 .11. 2. la, following any 20 
* 2.3 2.3.1          Boundary    Dose  Rate Salem ODCM GastOg* lffluent Instantaneous Dose Rate Calculations -
* *
10 CD 20 sit*                                 Noble   Gases.
* Salem ODCM Rev. a release exceeding the above prescribed alarm setpoints.
Rev. 8 Technical Specification 3.11.2.la limits the dose rate at the SITE BOUNDARY due to noble gas releases to     ~500    mrem/yr, total body and     ~3000 mrem/yr, skin. Radiation monitor alarm setpoints are established to ensure that these release limits are not exceeded.       In the event any gaseous releases from the station results in an alarm setpoint being exceeded,     an evaluation of the SITE BOUNDARY dose rate resulting from the release shall be performed using the following equations:
Monitor indications
( 2. 4) and D1 = X/Q
{readings) may be averaged over a time period not to exceed 15 minutes when determining noble gas release rate based on correlation of the monitor reading and monitor sensitivity.
* l: ( (~ + 1.1~)
The 15 minute averaging is needed to allow for reasonable monitor response to potentially changing radioactive material concentrations and to exclude potential electronic spikes in monitor readings that may be unrelated to radioactive material releases.
* Qi)   (2. 5) where:
As identified, any electronic spiking monitor responses may be excluded from the analysis.
            = total body dose rate (mrem/yr)                       .,
NOTE: For administrative purposes, more conservative alarm setpoints than those as prescribed above may be imposed. However, conditions exceeding these more limiting alarm setpoints do not necessarily indicate radioactive material release rates exceeding the limits of Technical Specification 3.11.2.la.
            = skin   dose rate (mrem/yr)                           _}
Provided actual releases do not result in radiation monitor indications exceeding alarm setpoint values based on the above criteria, no further analyses are required for demonstrating compliance with the limits of Specification 3.11.2.la. .l: Actual meteorological conditions concurrent with the period or the default, annual average dispersion parameters as presented in Table 2-4 may be used for evaluating the gaseous effluent dose rate. Sit* Boundary Dose Rate -Radioiodine and Particulates.
            = atmospheric dispersion to3 the controlling SITE BOUNDARY location (sec/m )
Technical Specification 3.11.2.1.b limits the dose rate to lives greater than 8 days. To demonstrate compliance with this 21 
            = average release rate of radionuclide i over               the release_ period under evaluation (uCi/sec)
* *
            = total body dose conversion factor             for noble   gas radionuclide i (mrem/yr per uci/m3 , from Table 2-1)
* Salem ODCM Rev. 8 limit, an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period (e.g., nominally once per 7 days). The following equation shall be used for the dose rate evaluation:
            = beta skin dose conversion factor             for noble     gas radionuclide i (mrem/yr per uCi/m3 , :fro~ Table 2-1)
where: X/Q = = (2. 6) average organ dose rate over the sampling time period (mrem/yr) atmospheric dispersion to the controlling*
            - q&J11Da air dose conversion factor for noble               gas radionuclide i (mrad/yr per uci/m3 , from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad)
SITE BOUNDARY location for the inhalation pathway (sec/m 3) dose parameter for radionuclide i (mrem/yr per uCi/m 3) and organ o for the child inhalation pathway from Table 2-5 average release rate over the appropriate sampling period and analysis frequency for radionuclide i I-131, I-133, ttitium or other radionuclide in particulate form with half-life greater than 8 days (uCi/sec)
As appropriate, simultaneous releases from sa_lem Units 1 and 2 and Hope creek will be considered in evaluating compliance with the releas*e rate limits of Specification 3 .11. 2. la, following any
By substituting 1500 mrem/yr for D 0 and solving for Qu an allowable  
* 20
., release rate for I-131 can be determined.
* Salem ODCM release exceeding the above prescribed alarm setpoints.
Based on th *. annual average meteorological dispersion (see Table 2-4) and the most limiting potential pathway, age group and organ (inhalation, child, thyroid --Rm= l.62E+07.mrem/yr per uCi/m 3), the allowable release *rate for I-131 is 42 uCi/sec. Reducing this:release rate by a factor of 4 to account for potential dose contributions from other radioactive particulate material and other release points (e.g., Hope Creek), the corresponding release rate allocated to each of 22 
Rev. a Monitor indications {readings) may be averaged over a time period not to exceed 15 minutes when determining noble gas release rate based on correlation of the monitor reading and monitor sensitivity. The 15 minute averaging is needed to allow for reasonable monitor response to potentially changing radioactive material concentrations and to exclude potential electronic spikes in monitor readings that may be unrelated to radioactive material releases.     As identified, any electronic spiking monitor responses may be excluded from the analysis.
* *
NOTE: For administrative purposes, more conservative alarm setpoints than those as prescribed above may be imposed.
* Salem ODCM Rev. s the Sal.em units is 10.5 uCi/sec. For a 7 day period, which is the nominal sampling and analysis frequency for I-131, the cumulative release is 6.3 Ci. Therefore, as long as the I-131 releases in any 7 day period do not exceed 6. 3 Ci, no additional analyses are needed for verifying compliance with the Technical Specification 3.11.2.1.b limits on allowable release rate
However, conditions exceeding these more limiting alarm setpoints do not necessarily indicate radioactive material release rates exceeding the limits of Technical Specification 3.11.2.la. Provided actual releases do not result in radiation monitor indications exceeding alarm setpoint values based on the above criteria, no further analyses are required for demonstrating compliance with the limits of Specification 3.11.2.la.
* 23
                                                            .l:
* *
Actual meteorological conditions concurrent with the relea~e period or the default, annual average dispersion parameters as presented in Table 2-4 may be used for evaluating the gaseous effluent dose rate.
* Salem ODCM Rev. 8 2.4 Noble Gaa Effluent Dose Calculations  
Sit* Boundary Dose Rate - Radioiodine and Particulates.
-10 CFR so trlfBISTRICTBD AREA Dose Noble Gases. Technical Specification 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate compliance with the quarterly*
Technical Specification 3.11.2.1.b limits the dose rate to     ~1500 lives greater than 8 days. To demonstrate compliance with this
dose limits of ss mrad, gamma-air and SlO mrad, beta-air and the calendar year limits SlO mrad, gamma-air and s20 mrad, beta-air.
* 21
The limits are applicable separately to each unit and are not combined site limits. The following equations shall be used to calculate the gamma-air and beta-air doses: where: = = X/Q = 3 .17E-08 = = 3.17E-08
* Salem ODCM   Rev. 8 limit, an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period (e.g.,
nominally once per 7 days). The following equation shall be used for the dose rate evaluation:
(2. 6) where:
average organ dose rate over the sampling time period (mrem/yr)
X/Q =  atmospheric dispersion to the controlling* SITE BOUNDARY location for the inhalation pathway (sec/m3 )
        ~o  =  dose parameter for radionuclide i (mrem/yr per uCi/m3 )
and organ o for the child inhalation pathway from Table 2-5 average release rate over the appropriate sampling period and analysis frequency for radionuclide i --
I-131, I-133, ttitium or other radionuclide in particulate form with half-life greater than 8 days (uCi/sec)
By substituting 1500 mrem/yr for D0 and solving for Qu an allowable release rate for I-131 can be determined.       Based on th*. annual average meteorological dispersion (see Table 2-4)       and the most limiting potential pathway, age group and organ (inhalation, child, thyroid -- Rm= l.62E+07.mrem/yr per uCi/m3 ), the allowable release
  *rate for I-131 is 42 uCi/sec. Reducing this:release rate by a factor of 4 to account for potential dose contributions from other radioactive particulate material and other release points (e.g.,
Hope Creek), the corresponding release rate allocated to each of
* 22
* the Sal.em units is 10.5 uCi/sec.
Salem ODCM Rev. s For a 7 day period, which is the nominal sampling and analysis frequency for I-131, the cumulative release is 6.3 Ci. Therefore, as long as the I-131 releases in any 7 day period do not exceed 6. 3 Ci, no additional analyses are needed for verifying   compliance with the Technical Specification 3.11.2.1.b limits on allowable release rate *
* 23
* 2.4 trlfBISTRICTBD    AREA  Dose Salem ODCM Noble Gaa Effluent Dose Calculations - 10 CFR so Noble Gases.
Rev. 8 Technical Specification 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate compliance with the quarterly* dose limits of     ss   mrad,   gamma-air and SlO mrad,   beta-air and the calendar year limits SlO mrad, gamma-air and s20 mrad, beta-air.
The limits are applicable separately to each unit and are not combined site limits.         The following equations shall be used to calculate the gamma-air and beta-air doses:
                      =   3.17E-08
* X/Q
* X/Q
* E
* E   {~ * ~)              (2.7) and
* and = 3.17E-08
* where:
            =
                      =   3.17E-08
* X/Q
* X/Q
* E
* E {~ * ~)
* air dose due to gamma emissions for noble g*(\ls radionuclides (mrad) ,,\: air dose due to beta emissions for noble gas radionuclides (mrad) (2.7) ( 2_. 8) atmospheric dispersion to the controlling SITE BOUNDARY location {sec/m3) cumulative release of noble gas radionuclide i over the period of interest {uCi) where uci = {uci/cc} (cc released) or (uCi/sec) (sec released) air dose factor due to gamma from noble gas radionuclide i (mrad/yr per uCi/m 3 , from Table 2-1) air dose factor due to beta emissions from noble gas radionuclide i (mrad/yr per uCi/m 3 , Table 2-1) conversion factor (yr/sec) 24 
air dose due to gamma emissions for noble g*(\ls
* *
( 2_. 8) radionuclides (mrad)                         ,,\:
* Salem ODCM Rev. 8 2.4.2 Sipplifie4 Dose Calculation for Noble Gases. In lieu of the individual noble gas radionuclide dose assessment as presented above, the following simplified dose calculation equations shall be used for verifying compliance with the dose limits of Technical Specification J.11.2.2. (Refer to Appendix c for the derivation and justification for this simplified method.) o, where: MetT = 0.50 = J.17E-08 = (2. 9) 0.50 and J.17E-08 =
            =      air dose due to beta emissions for noble gas radionuclides (mrad)
X/Q =      atmospheric dispersion to the controlling SITE BOUNDARY location {sec/m3) cumulative release of noble gas radionuclide i over the period of interest {uCi) where uci = {uci/cc} (cc released) or (uCi/sec) (sec released) air dose factor due to gamma emis~ions. from noble gas radionuclide i (mrad/yr per uCi/m3 , from Table 2-1) air dose factor due to beta emissions from noble gas radionuclide i (mrad/yr per uCi/m3 , Table 2-1) 3 .17E-08 =      conversion factor (yr/sec)
* 24
* 2.4.2   Sipplifie4 Dose Calculation for Noble Gases.
Salem ODCM individual noble gas radionuclide dose assessment as presented Rev. 8 In lieu of the above, the following simplified dose calculation equations shall be used for verifying compliance with the dose limits of Technical Specification J.11.2.2.       (Refer to Appendix c for the derivation and justification for this simplified method.)
J.17E-08 o,    =                                       (2. 9) 0.50 and J.17E-08
                        =
* X/Q
* X/Q
* Neff
* Neff
* l: Qi (2 .10) o. 50 v 5.JE+02, effective gamma-air dose factor (mrad/yr per uCi/m 3) 1. lE+OJ, effective beta-air dose factor ..(mrad/yr per uCi/m 3) : 'l cumulative release for all noble gas radibnuclides (uCi) where uCi = * (uCi/cc) (cc released) or (uCi/sec) (sec released) conservatism factor to account for potential variability in the radionuclide distribution Actµal meteorological conditions concurrent witQ the, release period or the default, annual average dispersion parameters as presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doses
* l: Qi     (2 .10)
* 25
: o. 50 v
* *
where:
* Salem ODCM Rev. a .2.5 Radioiodin*
MetT =      5.JE+02, effective gamma-air dose factor (mrad/yr per uCi/m3 )
and Particulate Dose Calculations  
: 1. lE+OJ, effective beta-air dose factor ..(mrad/yr per uCi/m3 )                                     : 'l cumulative release for all noble gas radibnuclides (uCi) where uCi = * (uCi/cc) (cc released) or (uCi/sec) (sec released) 0.50 =        conservatism factor to account for potential variability in the radionuclide distribution Actµal meteorological conditions concurrent witQ the, release period or the default, annual average dispersion parameters as presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doses *
-10 CFR so 2.s.1 tJNBBSTBICTED AREA Dose -Radioiodine and Particulates.
* 25
In accordance with reqllirements of Technical Specification 3 .11. 2. 3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit of mrem and calendar year limit mrem to any organ. The following equation shall be used to evaluate the maximum organ dose due to releases of I-131, tritium and particulates with half-lives greater than 8 days: Daop where: D = aop w = Ru,, = = = = 3. l 7E-08
 
Salem ODCM     Rev. a
* .2.5   Radioiodin* and Particulate Dose Calculations - 10 CFR so 2.s.1     tJNBBSTBICTED AREA Dose - Radioiodine and Particulates.
In   accordance         with     reqllirements   of   Technical     Specification 3 .11. 2. 3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit of               ~7.5 mrem and calendar year limit       ~15  mrem to any organ.     The following equation shall be used to evaluate the maximum organ dose due to releases of I-131, tritium and particulates with half-lives greater than 8 days:
Daop   =   3. l 7E-08
* W
* W
* SFp * :t (Rq,
* SFp * :t (Rq,
* Qi) (2.11) dose or dose commitment via all pathways p and controlling age group a {as identified in Table 2-4) to organ *O, including the total body {mrem) atmospheric dispersion parameter to the controlling location{s) as in Table 2-4 X/Q D/Q = atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways {sec/m3) * = atmospheric deposition for vegetation, milk and ground plane exposure pathways (m-2) dose factor for radionuclide i {mrem/yr per or {m 2 -mrem/yr per uCi/sec) and organ o from Table 2-s for each age group and the applicable pathway p as identified in Table 2-4. Values for Rm were derived in accordance with the methods described in NUREG-0133 cumulative release over the period of interest for radionuclide i --I-131 or radioactive material in particulate form with half-life than 8 days (uCi). -. annual seasonal correction factor to account for the fraction of the year that the applicable exposure pathway does not exist. 1) For milk and vegetation exposure pathways:  
* Qi)         (2.11) where:
= A six month fresh vegetation and grazing season. (May through October) = 0.5 2) For inhalation and ground plane exposure pathways:  
Daop =      dose or dose commitment via all pathways p and controlling age group a {as identified in Table 2-4) to organ *O, including the total body {mrem) w =          atmospheric dispersion parameter to the controlling location{s) as ide~tified in Table 2-4
= 1.0 26
* X/Q D/Q
* *
                              =
* Salem ODCM Rev. 8 For evaluating the maximum exposed individual, the infant age group is controlling for the milk pathway. Only the controlling age group as identified in Table 2-4 need be evaluated for compliance with Technical Specification 3.11.2.3.
                              =
2.s.2 simplified Dose Calculation for Radioiodines and Particulates.
atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways {sec/m3)
In lieu of the individual radionuclide (I-131 and particulates) dose assessment as presented above, the following simplified dose calculation equation may be used for verifying compliance with the dose limits of Technical Specification
* atmospheric deposition for vegetation, milk and ground plane exposure pathways (m-2 )
: 3. 11. 2. 3 (refer to Appendix D for the derivation and justification of this simplified method). where: = Dmu = 3 . 17E-08
Ru,, =        dose factor for radionuclide i {mrem/yr per ~i/m3 ) or
{m2 - mrem/yr per uCi/sec) and organ o from Table 2-s for each age group and the applicable pathway p as identified in Table 2-4. Values for Rm were derived in accordance with the methods described in NUREG-0133
                =      cumulative release over the period of interest for radionuclide i -- I-131 or radioactive material in particulate form with half-life g~eater than 8 days (uCi).                                   -   .
                =      annual seasonal correction factor to account for the fraction of the year that the applicable exposure pathway does not exist.
: 1) For milk and vegetation exposure pathways:
                              = A six month fresh vegetation and grazing season. (May through October)
                              =   0.5
: 2)   For inhalation and ground plane exposure pathways:
                              = 1.0 26
* Salem ODCM For evaluating the maximum exposed individual, the infant age group is controlling for the milk pathway.
Rev. 8 Only the controlling age group as identified in Table 2-4 need be evaluated for compliance with Technical Specification 3.11.2.3.
2.s.2     simplified         Dose   Calculation     for     Radioiodines       and Particulates. In lieu of the individual radionuclide (I-131 and particulates) dose assessment as presented above, the following simplified dose calculation equation may be used for verifying compliance with the dose limits of Technical Specification 3. 11. 2. 3 (refer to Appendix D for the derivation and justification of this simplified method).
Dmu   =   3 . 17E-08
* W
* W
* SFp
* SFp
* Rr.131
* Rr.131
* I: Qi ( 2. 12)
* I: Qi       ( 2. 12) where:
* maximum organ dose (mrem) I-131 dose parameter for the thyroid for the identified controlling pathway 1.05E+12, infant thyroid dose parameter with the cow-milk pathway controlling (m 2 -
maximum organ dose (mrem)
per uci/sec) ,Y D/Q for radioiodine, 2.lE-10 1/m 2
I-131 dose parameter for the thyroid for the identified controlling pathway
* cumulative release over the period of interest for radionuclide I --I-131 or radioactive material in particulate from with half life greater than a days (uCi) The location of exposure pathways and the maximum organ dose . :;i calculation may be based on the available pathways in the surrounding environment of Salem as identified by the annual use census (Technical Specification 3.12.2). Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4 . 27
            =    1.05E+12, infant thyroid dose parameter with the cow-milk pathway controlling (m2 - mrem/y~ per uci/sec)                                               ,Y D/Q for radioiodine, 2.lE-10 1/m2
* Salem ODCM Rev. 8 2.6 secondary Side Radioactive Gaseous Effluents and Dose Calculations During periods of primary.to secondary leakage, minor levels of radioactive material may be released via the secondary system to the atmosphere.
* cumulative release over the period of interest for radionuclide I -- I-131 or radioactive material in particulate from with half life greater than a days (uCi)
Non-condensables (e.g., noble gases) will be predominately released via the condenser evacuation system and will* be monitored and quantified by the routine plant vent monitoring and sampling system and procedures (e.g., RlS on condenser evacuation, R41C on plant vent, and the plant vent particulate and charcoal samplers).
The location of exposure pathways and the maximum organ dose
                                                                    . :;i calculation may     be     based   on   the available     pathways       in the surrounding environment of Salem as identified by the annual land-use census (Technical Specification 3.12.2).               Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4 .
* 27
* 2.6   secondary Calculations Side Radioactive Gaseous Salem ODCM Effluents Rev. 8 and   Dose During periods of primary.to secondary leakage, minor levels of radioactive material may be released via the secondary system to the atmosphere. Non-condensables (e.g., noble gases)     will be predominately released via the condenser evacuation system and will*
be monitored and quantified by the routine plant vent monitoring and sampling system   and procedures (e.g.,   RlS on condenser evacuation, R41C on plant vent, and the plant vent particulate and charcoal samplers).
However, if the Steam Generator blowdown is routed directly to the
However, if the Steam Generator blowdown is routed directly to the
* Chemical Waste Basin (via the SG blowdown flash tank) instead of being recycled through the condenser, it may be desirable to account for the potential atmospheric releases of radioiodines and particulates from the flash tank vent (i.e.,
* Chemical Waste Basin (via the SG blowdown flash tank) instead of being recycled through the condenser,     it may be desirable to account for the potential atmospheric releases of radioiodines and particulates from the flash tank vent       (i.e., release~; due to t:
due to ** t: moisture carry over)
moisture carry over)
* Since this pathway is not sampled or monitored, it is necessary to calculate potential releases.
* Since this pathway is not sampled or monitored, it is necessary to calculate potential releases.
Based on the guidance in NRC NUREG-0133, tl}e releases of the radioiodinas and particulates shall be calculated by the equation:
Based on the guidance in NRC NUREG-0133,     tl}e releases of the radioiodinas and particulates shall be calculated by the equation:
(2.13) where: Qi = the release rate of radionuclide, i, from the steam generator flash tank vent (uCi/sec) 28
(2.13) where:
* Ci = = Fft = SQ!tv = Salem ODCM Rev. 8 the concentration of radionuclide, i, in the secondary coolant water averaged over not more than one week (uCi/ml) the steam generator blowdown rate to the flash tank (ml/sec) the fraction of blowdown flashed in the tank determined from a heat balance taken around the flash tank at the applicable reactor power level
Qi   = the release rate of radionuclide, i, from the steam generator flash tank vent (uCi/sec)
* the measured steam quality in the flash tank vent; or an assumed value of 0.85, based on NUREG-0017.
**                                  28
Tritium releases via the steam flashing may also be quantified using the above equation with the assumption of a steam quality (SQ1tv) equal to o. Since the H-3 will be associated with the water molecules, it is not necessary to account for the moisture carry-over which is the transport media for the radioiodines and particulates.
* Ci
* Based on the design and operating conditions at Salem, the fraction of blowdown converted to steam (Fft) is approximately
      ~b      =
: o. 48. The simplifies to the following:  
                =
* (2 .14) For H-3, the simplified equation is: Also durinq reactor shutdown operations with a radioactively contaminated secondary system, radioactive material may be released to the atmosphere via the atmospheric reliefs (PORV) and the safety 29 
Salem ODCM     Rev. 8 the concentration of radionuclide, i, in the secondary coolant water averaged over not more than one week (uCi/ml) the steam generator blowdown rate to the flash tank (ml/sec)
* *
Fft    =    the fraction of blowdown flashed in the tank determined from a heat balance taken around the flash tank at the applicable reactor power level
* Salem ODCM Rev. a reliefs on the main steam lines and via the steam driven auxiliary feed pump exhaust. The evaluation of the radioactive material concentration in the steam relative to that in the steam generator water is based on the guidance of NUREG-0017, Revision 1. The partitioning factors for the radioiodines is 0.01 and is 0.001 for all other particulate radioactive material.
* SQ!tv  =    the measured steam quality in the flash tank vent; or an assumed value of 0.85, based on NUREG-0017.
The resulting equation for quantifying releases via the atmospheric steam releases is: where: Qij = cij = SFj = = = = PFi = = = = 0.13= (2 .16) release rate of radionuclide i via pathway j {uCi/sec) concentration of radionuclide i, in pathway j, (uCi/sec) steam flow for.release pathway j 450,000 lb/hr per PORV . 800,000 lb/hr per safety relief valve 50,000 lb/hr for auxiliary feed pump exhaust partitioning factor, ratio of concentration in steam to that in the water in the steam generator 0.01 for radioiodines 0.005 for all other particulates
Tritium releases via the steam flashing may also be quantified using the above equation with the assumption of a steam quality (SQ1tv) equal to   o. Since the H-3 will be associated with the water molecules, it is not necessary to account for the moisture carry-over which       is the   transport media for the   radioiodines   and particulates.
: 1. o for H-3 conversion factor -[(hr*ml) / (sec*lb)]
* Based on the design and operating conditions at Salem, the fraction of blowdown converted to steam (Fft) is approximately o. 48.           The equa~ion      simplifies to the following:
Any significant releases of noble gases via the atmospheric steam releases can be quantified in accordance with the calculation methods the Salem Emergency Plan Implementation*
(2 .14)
Procedure . 30
For H-3, the simplified equation is:
-* *
Also     durinq reactor shutdown operations with       a radioactively contaminated secondary system, radioactive material may be released to the atmosphere via the atmospheric reliefs (PORV) and the safety
* Salem ODCM Rev. 8 Alternately, the quantification of the release rate and cumulative releases may be based on actual samples of main steam collected at the R4 6 sample locations.
* 29
The measured radionuclide concentration in the steam may be used for quantifyinq the noble gases, radioiodine and particulate releases.
* Salem ODCM     Rev. a reliefs on the main steam lines and via the steam driven auxiliary feed pump exhaust. The evaluation of the radioactive material concentration in the steam relative to that in the steam generator water is based on the guidance of NUREG-0017, Revision 1.         The partitioning factors for the radioiodines is 0.01 and is 0.001 for all other particulate radioactive material. The resulting equation for quantifying releases via the atmospheric steam releases is:
Note: The expected mode of operation would be to isolate the effected steam generator, thereby reducinq the potential releases durinq the shutdown/cooldown process. Use of the above calculation methods should consider actual operating conditions and release mechanisms.
(2 .16) where:
The calculated quantities of radioactive materials may be used as inputs to the equation (2.11) or (2.12) to calculate offsite doses for demonstratinq compliance with the Radioloqical Effluent Technical Specifications
Qij =  release rate of radionuclide i via pathway j {uCi/sec) cij =  concentration of radionuclide i, in pathway j, (uCi/sec)
SFj  =  steam flow for.release pathway j
          =  450,000 lb/hr per PORV           .
          =  800,000 lb/hr per safety relief valve
          =  50,000 lb/hr for auxiliary feed pump exhaust PFi =  partitioning factor, ratio of concentration in steam to that in the water in the steam generator
          =  0.01 for radioiodines
          =  0.005 for all other particulates
          =  1. o for H-3 0.13=  conversion factor - [(hr*ml) / (sec*lb)]
Any significant releases of noble gases via the atmospheric steam releases can be quantified in accordance with the calculation methods o~ the Salem Emergency Plan Implementation* Procedure .
* 30
 
-*                                               Salem ODCM Rev. 8 Alternately, the quantification of the release rate and cumulative releases may be based on actual samples of main steam collected at the R4 6 sample locations. The measured radionuclide concentration in the   steam may be used   for quantifyinq the noble gases, radioiodine and particulate releases.
Note:   The expected mode of operation would be to isolate the effected steam generator, thereby reducinq the potential releases durinq the shutdown/cooldown process. Use of the above calculation methods should consider actual operating conditions and release mechanisms.
The calculated quantities of radioactive materials may be used as inputs to the equation (2.11) or (2.12) to calculate offsite doses for   demonstratinq compliance   with the Radioloqical   Effluent Technical Specifications *
* 31
* 31
* Salem ODCM Rev. 8 2.7 Gaseou* Bffluent Dose Proiection Technical Specification 3 .11. 2. 4 requires that the GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when projected doses exceed one-half the annual design objective rate in any calendar quarter, i.e., exceeding:
* 2.7 Gaseou* Bffluent Dose Proiection Salem ODCM    Rev. 8 Technical Specification 3 .11. 2. 4 requires that the GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when projected doses exceed one-half the annual design objective rate in any calendar quarter, i.e., exceeding:
0.625 mrad/quarter, gamma air; 1.25 mrad/quarter, beta air; or 1.875 mrem/quarter, maximum organ. The applicable gaseous processing systems for maintaining radioactive material releases ALARA are the Auxiliary Building normal ventilation system (filtration systems # 1,2 and 3) and the Waste Gas Decay Tanks as delineated in Figures 2-3 and 2-4 * ..
0.625 mrad/quarter, gamma air; 1.25 mrad/quarter, beta air; or 1.875 mrem/quarter, maximum organ.
The   applicable   gaseous processing     systems   for   maintaining radioactive material releases ALARA are the Auxiliary Building normal ventilation system (filtration systems # 1,2 and 3) and the Waste Gas Decay Tanks as delineated in Figures 2-3 and 2-4 *
* Dose projections are performed at least once per 31 days by. the following equations:
* Dose projections are performed at least once per 31 days by. the following equations:
where: DIP = D, = Dtip ---Dmup = Dmu = d = 91 = ** DIP = o, * (91 I d) (2.17) " Dtip = Db * (91 I d) (2 *:Ja> Dmap = Dmu * (91 I d) (2 .19) gamma air dose projection for current calendar quarter (mrad) gamma. air dose to date for current calendar quarter as determined by Equation 2.7 or (mrem) beta air dose projection for current calendar quarter (mrad) beta air dose to date for current calendar quarter as determined by Equation 2.8 or 2.10 (mrem) maximum organ dose projection for current calendar quarter (mrem) maximum organ dose to date for current calendar quarter as determined by Equation 2.11 or 2.12 (mrem) number of days to date in current calendar quarter number of days in a calendar quarter 32
DIP   = o, * (91 I d)                 (2.17)
* * ** Salem ODCM Rev. a 3.0 Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY In accordance with Technical Specification
Dtip   = Db * (91 I d)                 (2 *:Ja>
: 6. 9 .1.11, the Radioactive Effluent Release Report (RERR) submitted within 60 days after January 1 of each year shall include an assessm,ent of radiation doses from radioactive liquid and qaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.
Dmap   = Dmu * (91 I d)                 (2 .19) where:
DIP  =  gamma air dose projection for current calendar quarter (mrad)
D,  =  gamma. air dose to date for current calendar quarter as determined by Equation 2.7 or ~.9 (mrem)
Dtip -- beta air dose projection for current calendar quarter
        ~    -
Dmup =
(mrad) beta air dose to date for current calendar quarter as determined by Equation 2.8 or 2.10 (mrem) maximum organ dose projection for current calendar quarter (mrem)
Dmu  =  maximum organ dose to date for current calendar quarter as determined by Equation 2.11 or 2.12 (mrem) d    =  number of days to date in current calendar quarter
**      91  =  number of days in a calendar quarter 32
* 3.0 3.1 Special Dose Analyses Salem ODCM Doses Due To Activities Inside the SITE BOUNDARY Rev. a In   accordance   with   Technical Specification   6. 9 .1.11,   the Radioactive Effluent Release Report (RERR) submitted within 60 days after January 1 of each year shall include an assessm,ent of radiation doses from radioactive liquid and qaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.
The calculation methods as presented in Sections 2.4 and 2.5 may be used for determining the maximum potential dose to a MEMBER OF THE PUBLIC based on the parameters from Table 2-4
The calculation methods as presented in Sections 2.4 and 2.5 may be used for determining the maximum potential dose to a MEMBER OF THE PUBLIC based on the parameters from Table 2-4
* The default value for the meteorological dispersion data as presented in Table 2-3 may be used if current year meteorology is unavailable at .;the time .. l-of NRC reporting.
* The default value for the meteorological dispersion data as presented in Table 2-3 may be used if current year meteorology is unavailable at .;the time
However, a follow-up evaluation  
                                                              . l-of NRC reporting.     However, a follow-up evaluation 'shall     be performed when the data becomes available *
'shall be performed when the data becomes available
**                                33
* 33
* 3.2 Salem ODCM Total dose to MEMBERS OP THE PUBLIC - 40 CJ'R 190 Rev. a The Radioactive Effluent Release Report (RERR) submitted within 60 days after January 1 of each year shall also include an assessment of the radiation dose to the likely most exposed MEMBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle sources (including dose contributions from. effluents and direct radiation from on-site sources). For the likely most exposed MEMBER OF THE PUBLIC in the vicinity of Artificial Island, the sources of exposure need only consider the Salem Nuclear Generating Station and the Hope creek Nuclear Generating Station:     No other
* Salem ODCM Rev. a 3.2 Total dose to MEMBERS OP THE PUBLIC -40 CJ'R 190 The Radioactive Effluent Release Report (RERR) submitted within 60 days after January 1 of each year shall also include an assessment of the radiation dose to the likely most exposed MEMBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle sources (including dose contributions from. effluents and direct radiation from on-site sources).
* fuel cycle facilities contribute to the MEMBER OF THE PUBLIC dose for the Artificial Island vicinity.
For the likely most exposed MEMBER OF THE PUBLIC in the vicinity of Artificial Island, the sources of exposure need only consider the Salem Nuclear Generating Station and the Hope creek Nuclear Generating Station: No other
The dose contribution from the operation of Hope Creek,}Nuclear Generating Station will be estimated based on the methods as presented in the Hope Creek Offsite Dose Calculation Manual (HCGS ODCM).
* fuel cycle facilities contribute to the MEMBER OF THE PUBLIC dose for the Artificial Island vicinity.  
As appropriate for demonstrating/evaluating compliance with the limits of Technical Specification 3.11.4 (40 CFR 190), the results of the environmental monitoring program may be used for providing data on actual measured levels of radioactive material in the actual pathways of exposure .
* " The dose contribution from the operation of Hope Creek,}Nuclear Generating Station will be estimated based on the methods as presented in the Hope Creek Offsite Dose Calculation Manual (HCGS ODCM). As appropriate for demonstrating/evaluating compliance with the limits of Technical Specification 3.11.4 (40 CFR 190), the results of the environmental monitoring program may be used for providing data on actual measured levels of radioactive material in the actual pathways of exposure . 34
* 34
* Salem ODCM Rev. a 3. 2 .1 Bfflutnt Dost Calculations.
* 3. 2 .1 Bfflutnt Dost Calculations.
For purposes of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.11 (RERR), dose calculations for the Salem Nuclear Generating Station may be performed using the calculation methods contained within this .ODCM; the conservative controlling pathways and locations of Table 2-4 or the actual pathways and locations as identified by the land use census (Technical Specification 3/4.12.2) may be used. Average annual meteorological dispersion parameters or meteorological conditions concurrent with the release period under evaluation may be used. l.2.2 Direct Exposure Dose Determination.
Salem ODCM  Rev. a For purposes of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.11 (RERR), dose calculations for the Salem Nuclear Generating Station may be performed using the calculation methods contained within this .ODCM; the conservative controlling pathways and locations of Table 2-4 or the actual pathways   and locations as   identified by the   land use census (Technical Specification 3/4.12.2) may be used.       Average annual meteorological dispersion parameters or meteorological conditions concurrent with the release period under evaluation may be used.
Any potentially
l.2.2   Direct   Exposure   Dose   Determination. Any   potentially
* significant direct exposure contribution to. off-site individual doses may be evaluated based on the results of the environmental
* significant direct exposure contribution to. off-site individual doses may be evaluated based on the results of the environmental measurements (e.g., TLD, ion chamber measurements) and/or by the use of a radiation transport and shielding calculation} method.
* measurements (e.g., TLD, ion chamber measurements) and/or by the ,, use of a radiation transport and shielding calculation}
Only during atypical conditions will there exist any potential for significant on-site sources at Salem that wouid yield potentially significant off-site doses (i.e., in excess of 1 mrem per year to a MEMBER OP THE PUBLIC) , that would require detailed evaluation for demon*stratinC) compliance with 40 CFR 190.       However,   should a situation   exist whereby   the direct exposure   contribution     is potentially     significant,     on-site   measurements,       off-site measurements   and/or calculation   techniques   will be used     for determination of dose for assessing 40 CFR 190 compliance .
method. Only during atypical conditions will there exist any potential for significant on-site sources at Salem that wouid yield potentially significant off-site doses (i.e., in excess of 1 mrem per year to a MEMBER OP THE PUBLIC) , that would require detailed evaluation for demon*stratinC) compliance with 40 CFR 190. However, should a situation exist whereby the direct exposure contribution is potentially significant, on-site measurements, off-site measurements and/or calculation techniques will be used for determination of dose for assessing 40 CFR 190 compliance . 35 
* 35
* *
* 4.0 Salem ODCM Radiological Environmental Monitoring Program Rev. 8 4.1 Sampling Program The operational phase of the Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of Appendix A Technical Specification 3.12. The objectives of the program are:
* Salem ODCM Rev. 8 4.0 Radiological Environmental Monitoring Program 4.1 Sampling Program The operational phase of the Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of Appendix A Technical Specification 3.12. The objectives of the program are: -To determine whether any significant increases occur in the concentration of radionuclides in the critical pathways of exposure in the vicinity of Artificial Island; To determine if the operation of the Salem Nuclear Generating Stations has resulted in any increase in the inventory of long lived radionuclides in the environment; To detect any changes in the ambient gamma radiation levels; and To verify that SNGS operations have no detrimental effects on the health and safety of the public or on the The sampling requirements (type of samples*, collection frequency and analysis) and sample locations are presented in Appendix E. *uoTB: No public drinking water samples or irrigation water samples are taken as these pathways are not directly effected by liquid effluents discharged from Salem Generating Station
        - To determine whether any significant increases occur in the concentration of radionuclides in the critical pathways of exposure in the vicinity of Artificial Island; To determine if the operation of the Salem Nuclear Generating Stations has resulted in any increase in the inventory of long lived radionuclides in the environment;
* 36
* To detect any changes in the ambient gamma radiation levels; and To verify that SNGS operations have no detrimental effects on the health and safety of the public or on the environmen~.
* *
The sampling requirements (type of samples*, collection frequency and analysis) and sample locations are presented in Appendix E.
* Salem ODCM Rev. a 4.2 InterlaJ2oratory comparison Program Technical Specification 3.12.3 requires analyses be performed on radioactive material supplied as part of an Interlaboratory Comparison.
  *uoTB: No public drinking water samples or irrigation water samples are taken as these pathways are not directly effected by liquid effluents discharged from Salem Generating Station *
Participation in an approved Interlaboratory Comparison Program provides a check on the preciseness of measurements of radioactive materials in environmental samples. A summary of the Interlaboratory Comparison Proqram results.will be provided in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.10
* 36
* 37 ----------
* 4.2 InterlaJ2oratory comparison Program Salem ODCM  Rev.
w *
Technical Specification 3.12.3 requires analyses be performed on a
* u*-* . I *1 i::, i * '*; ... dBftfu .. -*-----'1 ....,_ '--=wt '--111(111' J"" ..... . 1.,... -m .... -----.... r IDI " ...
radioactive   material   supplied   as part   of an Interlaboratory Comparison.     Participation   in   an   approved   Interlaboratory Comparison   Program   provides   a   check on   the preciseness of measurements of radioactive materials in environmental samples.     A summary of the Interlaboratory Comparison Proqram results.will be provided in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.10 *
* RAOIATION MONITORING LIQUID RELEASES UNIT l FIGURE 1-1 c&£L J .. ........ ".:Jlfl& ... .,... lo--.--"IL"ll-....... r &.1*r r--.. .... Ilk ..,,.. m-.. CllllUIMlll 111P1 1111111 _J .... .... ..  
* 37
.: . . *. .. , ..... *-I* Lr . . .,.,. '* auau:aa *****---.. * ..
 
* I *' I*
RAOIATION MONITORING LIQUID RELEASES UNIT l FIGURE 1-1 u*-*                                             c&£L
w ..
                            .I   *~  *1 i::,
* u,_ .. dl&Ma .-----. ........ _ mlll'\.. .... *tm .._ Bl"' . ,_,.. -m .... -----*r..1, 1 * *r ... .... r
J. .. ..... Lr ..,.,.
* P* ***
* I
* RADIATION LIQUID RELEASES UNIT 2 FIGURE 1-2 . I J' er *.IL.'WI. * :JI. Ha ...... m ..... -.... _ .. -u&a'Tll*
:~ i   * '*; ...
I ....... r *I .. .... , r---& .. -I* I* ** *a !Ml' -** ... lf**-**11 m-.. .. Cl&llJlllllll
                          ..~
.... llllP _J ' -.... . . _.,,_: :*:*
                                ~-                                           ........ .-".:Jlfl&                  ,
* 0 * *llOTE: Evaporator p.cltage and/or 1'.aMste delltneraltierr s1ste11 ---IU , .............. . -*---*-** .. _ (lllC)
dBftfu
* I I I I I : I 1'0 ***** ..... -*-... ..... .... c .................. . --a"'"" rt *11, . .... **ac.* COO&.* ... --* ... -AC&:l .. IULAl'Ull U11&1*a **r.:a .. U*IMl*M ftg 1-l S.I*
            -*-----                         '1 lo--
* a.....u.a
                                                                                            "IL"ll-w
""""'-***
    -=wt  '--  111(111' J""      ..... r . . . . r                            &.1*r    r-- ...... Ilk        I*              '*          I*
W.ar i;,., .. .., FOi llFCNllATIOI aw 
  . 1.,...
* *
  -m m-IDI
* Salem ODCM Rev. 8 Table 1-1 Parameters for Liquid Alarm Setpoint Determinations Unit 1 Parameter Actwl Defailt ec-nta Value Value MPC 0 calculated 4.71E-06
                                                ..."      .. CllllUIMlll 111P1 1111111
* uci/ml calculated for each batch to be released.
_J                              *****---..
MPC I-131 3.0E-07 N/A uci/11l I-131 MPC conservatively used for SG blowdown and Service Water monitor set....,ints.
auau:aa
c, measured N/A uci/ml taken froia g1111111a spectral analysis of l fm.iid effluent.
                                                        . ~,-:._...;
14P(1 as det emi necl N/A uci/11l taken froia 10 CFR 20, Appendix B, Table II. Col 2. Sen 1-R18 as determined 2.9E+07 Cplll per Radwaate Effluent (Cs-137) uci/*l 1-R19 2.9E+07 Ste .. Generator
                                                              . *.~ *.
&lowdown ccs-137) (A,B,C,D) 1-R13 (A) 5.62E+07 Service Water -Contai1111ent Fan 1-R13 CB) 5.98E+07 Cooling ccs-137) 1-R13 (C,D,E) 1.01E+08 cw BS deten1inecl 1.00E+OS 9PI circulating
* RADIATION      ~ONITORING    ***  LIQUID RELEASES UNIT 2 FIGURE 1-2 u,_ ..                                           ~&.
... ter *sys.tern
I dl&Ma
-single cw m..a ., RR 1-R18 as detenained 120 aim detenained prior to release rate can be ad-justed for 1-R19 Technical Specification 1-R13 120 Ste .. Generator blowdown rate per Generator Service Water flow rate for 2500 Contai1111ent fan coolers Setpoint 1-R11 calculated 1.13E+05(+bkg)
                                        *r..1, 1 ~
Cplll Default alana setpoints:
                                                                            *.IL.'WI.   ~
more conser:vative*velues may be used as 1-R19 -7.25E+03C+blcg) deenied appropriate and desiralbe 1-R13 CA) .. for ensuring regulatory 6.70E+02(+bkg) and for .. intaining releases ALARA. 1-R13 (B) ..., 7.10E+02(+blcg) 1-R13 (C,D,E)-1.09E+03(+bkg)
                                                                                            ....* :JI._.. ...... J' Ha m -~ er .....
* Refer to Appendix .A for derivation  
                                                  *r mlll'\.. ....                                                                           -u&a'Tll*
** The MPC*value of I-131 C3E-07 uci/ml) has been used for derivation of R19 Ste .. Generator blowdown and R13 Service Water monitor setpoints as discussed in Section 1.Z.Z 41 I 
w
* *
  .-m,_,..
* Salem ODCM Rev. 8 Table 1-2 Parameters for Liquid Alarm setpoint Determinations Unit 2 Parameter Actual Default *units Conmen ts Value Value MPC 0 calculated 3.38E-06
    ~      .._
* uci/111l calculated for each batch to be released.
Bl"'
MPC 1*131 3.0E-07 N/A uci/ml 1*131 MPC conservatively used for SG blowdown, Service Water and Chemical Waste Basin monitor setnoints.
                                    *tm
: c. measured N/A uci/11l tak1n frCllll 98111111!
                                            .... . r . . . . r ....,                    r-- -&..-
spectral analysis of ll,.*id effluent.
I* I* **                          *I
MPC. as determined N/A uci /Ill taken frOll 10 CFR 20, Appendix B, Table II -Col. 2 Sen 2*R18 as determined 1.14E+08 cpnt/uc i /111l Redweate Effluent ccs-137) 2*R19CA,C) 1 .26E+08 Ste .. Generator Blowdown (Cs-137) 2*R19(B) 1. 14E+08 2*R19(D) 1. 13E+08 2*R13 9.05E+07 Service Water
                                                                                                                                          *a I
* Containnent Fan Cooling (Ca-137> .,_ R37 1.24E+08 C!'e11ical Waste Basin discharge cw as determined 1.0E+05 gpm Circulating Water System, single CW (Note: no CW in service for 2R13 monitor* see section 1.2.2) RR 2*R18 aa deten1i necl 120 9fm deten1inecl prior to release rate can be adjusted.:lr Technical Specification Ccq:il i *: e
    .... -----
* 2*R19 120 Ste .. Generator Slowdown rate per Generator Service Weter flow rate for 2*R13 2500 Contairaent fan coolers R37 1200 Chemical Waste Baain discharge Setpoint 2-R1a calculated 3.20E5 C+bkg) Cpnl Default alarm setpoints:
* P*
more conservative values may be used as 2-R19 -2. 10E4 deellld appropriate and desirable for
m-..
* CA,B,C,aJ ensuring regulatory ccq:iliance and for .. intaining releases ALARA. 2*R13 " 3.05E2 (+bkg) R37" 3. 10E3 (+bkg)
                                                                                                          !Ml' lf**-**11
* Refer to Appendix A for derivation
_J '
** The MPC value of 1*131 (3.0E-7 ucilml) has been used for derivation of the R13 and R37 monitor setpoints as discussed in Section 1.2.2 *** 2R19A setpoint calc -SC-RM-002-08, 2R19B setpoint calc -SC-RM-002*09, 2R19C setpoint calc -SC*RM-002-10, 2R19D setpoint calc
                                                            .. Cl&llJlllllll .... llllP
 
  *llOTE: Evaporator p.cltage and/or 1'.aMste delltneraltierr s1ste11 1'0 *****
AC&:l..IULAl'Ull U11&1*a
                                                                                                                      **r.:a.. U*IMl*M (lllC) 0 I
I I
I I                                        --*... -
                                                                                          **ac.* COO&.* ...
              --- -*--                        :I              --
                                                                .... c ...................
ftg 1-l S.I*
a.....u.a """"'-***     W.ar  i;,.,....,
IU                                                                                FOi llFCNllATIOI            aw a"'"" rt *11,
* Table 1-1 Salem ODCM              Rev. 8 Parameters for Liquid Alarm Setpoint Determinations Unit 1 Parameter              Actwl            Defailt          ~it*                    ec-nta Value            Value MPC 0              calculated        4.71E-06
* uci/ml      calculated for each batch to      be released.
MPC I-131              3.0E-07                N/A          uci/11l    I-131 MPC conservatively used for SG blowdown and Service Water monitor set....,ints.
c,              measured              N/A          uci/ml      taken froia g1111111a spectral analysis of l fm.iid effluent.
14P(1          as det emi necl          N/A          uci/11l    taken froia 10 CFR 20, Appendix B, Table II. Col 2.
Sen      1-R18      as determined        2.9E+07          Cplll per  Radwaate Effluent (Cs-137) uci/*l 1-R19                            2.9E+07                      Ste.. Generator &lowdown ccs-137)
(A,B,C,D) 1-R13 (A) 1-R13 CB) 1-R13 (C,D,E) 5.62E+07 5.98E+07 1.01E+08 Service Water - Contai1111ent Fan Cooling ccs-137)
I cw            BS  deten1inecl      1.00E+OS          9PI        circulating ...ter *sys.tern - single cw m..a                    .,
RR        1-R18    as detenained            120          aim        detenained prior to      n;l~ase:
release rate can be ad-justed for 1-R19                                                        Technical Specification c~liance 1-R13                              120                      Ste.. Generator blowdown rate per Generator Service Water flow rate for 2500                      Contai1111ent fan coolers Setpoint 1-R11        calculated      1.13E+05(+bkg)        Cplll      Default alana setpoints: more conser:vative*velues may be used as 1-R19 -                   7.25E+03C+blcg)                deenied appropriate and desiralbe for ensuring regulatory c~liance 1-R13 CA) ..                       6.70E+02(+bkg)                  and for .. intaining releases ALARA.
1-R13 (B) ...,                    7.10E+02(+blcg) 1-R13 (C,D,E)-                      1.09E+03(+bkg)
* Refer to Appendix .A for derivation
  ** The MPC*value of I-131 C3E-07 uci/ml) has    been used for derivation of R19 Ste.. Generator blowdown and R13 Service Water monitor setpoints as discussed in Section 1.Z.Z 41
* Table 1-2 Parameters for Liquid Alarm setpoint Determinations Unit 2 Salem ODCM                Rev. 8 Parameter              Actual            Default          *units                  Conmen ts Value              Value MPC0            calculated          3.38E-06
* uci/111l      calculated for each batch to be released.
MPC 1*131            3.0E-07              N/A         uci/ml        1*131 MPC conservatively used for SG blowdown, Service Water and Chemical Waste Basin monitor setnoints.
: c.               measured              N/A          uci/11l      tak1n frCllll 98111111! spectral analysis of ll,.*id effluent.
MPC.             as determined            N/A         uci /Ill      taken frOll 10 CFR 20, Appendix B, Table II - Col. 2 Sen        2*R18    as determined        1.14E+08    cpnt/uc i /111l  Redweate Effluent ccs-137) 2*R19CA,C)                           1.26E+08                       Ste.. Generator Blowdown (Cs-137) 2*R19(B)                      1. 14E+08 2*R19(D)                      1. 13E+08 2*R13                          9.05E+07                      Service Water
* Containnent Fan Cooling (Ca-137>
            .,_  R37                          1.24E+08                      C!'e11ical Waste Basin discharge cw              as determined        1.0E+05        gpm          Circulating Water System, single CW
                                                                            ~ (Note: no CW ~ in service for 2R13 monitor* see section 1.2.2)
RR        2*R18  aa deten1i necl          120        9fm          deten1inecl prior to r~ease: release rate can be adjusted.:lr Technical Specification Ccq:il i *: e
* 2*R19                           120                      Ste.. Generator Slowdown rate per Generator Service Weter flow rate for 2*R13                          2500                      Contairaent fan coolers R37                          1200                      Chemical Waste Baain discharge Setpoint 2-R1a        calculated      3.20E5 C+bkg)       Cpnl          Default alarm setpoints: more conservative values may be used as 2-R19 -                        2. 10E4                          deellld appropriate and desirable for
* CA,B,C,aJ                                                        ensuring regulatory ccq:iliance and for
                                                                            ..intaining releases ALARA.
2*R13 "                    3.05E2 (+bkg)
R37"                    3. 10E3 (+bkg)
* Refer to Appendix A for derivation
  ** The MPC value of 1*131 (3.0E-7 ucilml) has been used for derivation of the R13         and   R37 monitor setpoints as discussed in Section 1.2.2
  *** 2R19A setpoint calc - SC-RM-002-08, 2R19B setpoint calc - SC-RM-002*09, 2R19C setpoint calc - SC*RM-002-10, 2R19D setpoint calc
* SC*RM-002-11.
* SC*RM-002-11.
42 Salem ODCM Rev. 8 Table 1-3 (cont'd)
42
* Site Related Ingestion Dose Commitment Factors, Aio (mrem/hr per uCi/ml) Nuclide Bone Liver T.Body Thyroid Kidney Lung GI-LLI -------------------------------------------------
 
Salem ODCM    Rev. 8 Table 1-3 (cont'd)
Site Related Ingestion Dose Commitment Factors, Aio (mrem/hr per uCi/ml)
Nuclide    Bone      Liver  T.Body    Thyroid  Kidney      Lung      GI-LLI Ru-103    1. 07E+2            4.60E+l 4.07E+2
: 1. 25E+4 Ru-105    8.89E+O            3.SlE+O            l.15E+2                5.44E+3 Ru-106    l.59E+3          *2.01E+2              3.06E+3                1.03E+5 Rh-103m Rh-106 Ag-llOm    1. 56E+3  l.45E+3  8.60E+2              2.85E+3                5.91E+5 Sb-124    2.77E+
* 63
* 63
* APPENDIX A Evaluation of Default MPC Value for Liquid Effluents  
 
* **
APPENDIX A Evaluation of Default MPC Value for Liquid Effluents
* Salem ODCM Rev. 8 . Appendix A Evaluation of Default MPC Value for Liquid Effluents In accordance with the requirements of Technical Specification
* Salem ODCM   Rev. 8
{3.3.3.8) the radioactive liquid effluent monitors shall be operable with alarm setpoints established to ensure that the concentration of radioactive
                        .       Appendix A Evaluation of Default MPC Value for Liquid Effluents In accordance with the requirements of Technical Specification {3.3.3.8) the radioactive liquid effluent monitors shall be operable with alarm setpoints established to ensure that the concentration of radioactive
* material at the discharge point does not exceed the MPC value of 10 CFR 20, Appendix B, Table II, Column 2. The determination of allowable radionuclide concentration and correspondinq alarm setpoint is a function of the individual radionuclide distribution and corresponding
* material at the discharge point does not exceed the MPC value of 10 CFR 20, Appendix B, Table II, Column 2.     The determination of allowable radionuclide concentration   and correspondinq   alarm setpoint is a function of the individual radionuclide distribution and corresponding
* MPC values.
* MPC values.
* In order to limit the need for routinely havinq reestablish the alarm setpoints as a function of changinq radionuclide a default alarm setpoint can be established.
In order to limit the need for routinely havinq t~  reestablish the alarm setpoints as a   function of changinq radionuclide     di~,hibutions,  a default alarm setpoint can be established. This default setpoint can be based on an evaluation of the radionuclide distribution of the liquid effluents from Salem and the effective MPC value for this distribution.
This default setpoint can be based on an evaluation of the radionuclide distribution of the liquid effluents from Salem and the effective MPC value for this distribution.
The effective MPC value for a radionuclide distribution is calculated by the equation:
The effective MPC value for a radionuclide distribution is calculated by the equation:
A-2
* A-2
* Salem ODCM Rev. 8 E C 1 (qamma emitters only) MPCe = -----------------------------(A. l) Ci (qamma) Ci (non-qamma).
* Salem ODCM Rev. 8 E C1 (qamma emitters only)
E -----------
MPCe = -----------------------------
+ ---------------
Ci (qamma)       Ci (non-qamma).
where: MPCe = an effective MPC value for a mixture of radionuclide (uCi/ml) Ci = concentration of radionuclide i in the mixture MPCi = the 10 CFR 20, Appendix.
(A. l)
B, Table II, Column 2 MPC value for radionuclide i (uCi/ml) The equation for determining the liquid effluent setpoints ( Section 1.2.1, equation 1.2 ) is on a multiplication of the effective MPC
E -----------   + ~ ---------------
* times the monitor sensitivity.
where:
However, the *radiation monitors on the effluent lines will not detect non-gamma emitting radionuclides, such as
MPCe =   an effective MPC value for a mixture of radionuclide (uCi/ml)
* H-3, Fe-SS, and sr-90. The derivation of the effective_MPC ( section " 1.2.1, equation 1.3 ) is valid for any distribution but be modified to account for the fact that the effluent monitor will not detect the non-gammas.
Ci   = concentration of radionuclide i in the mixture MPCi =   the 10 CFR 20, Appendix. B, Table II, Column 2 MPC value for radionuclide i (uCi/ml)
The above modified equation for the effective MPC provides for a default setpoint determination that accounts for the non-gamma emitting radionuclides
The equation for determining the liquid effluent setpoints ( Section 1.2.1, equation 1.2 ) is   ba~d  on a multiplication of the effective MPC
* A-3
* times the monitor sensitivity. However, the *radiation monitors on the effluent lines will not detect non-gamma emitting radionuclides, such as H-3, Fe-SS, and sr-90. The derivation of the effective_MPC ( section 1.2.1, equation 1.3 ) is valid for any distribution but mu~t be modified to account for the fact that the effluent monitor will not detect the non-gammas. The above modified equation for the effective MPC provides for a default setpoint determination that accounts for the non-gamma emitting radionuclides *
* *
* A-3
* Salem ODCM Rev. 8 Considering the average effective MPC value for the years 1988 through 1990, it is reasonable to select an MPCe value of 4.71E-06 uCi/ml for Unit 1 and 3.38E-06 uci/ml for Unit 2 as tYi:>ical of liquid radwaste discharges.
* Salem ODCM Rev. 8 Considering the average effective MPC value for the years 1988 through 1990, it is reasonable to select an MPCe value of 4.71E-06 uCi/ml for Unit 1 and 3.38E-06 uci/ml for Unit 2 as tYi:>ical of liquid radwaste discharges. Using these values to calculate the default Rl8 alarm setpoint value, results in a setpoint that:
Using these values to calculate the default Rl8 alarm setpoint value, results in a setpoint that: 1) Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations, and 2) Will provide for a liquid radwaste discharge rate (as evaluated for each batch release) that is compatible with plant operations (refer to Tables 1-1 and 1-2) . A-4
: 1)   Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations, and
* Salem ODCM Rev. 8 Table A-1 Calculation of Effective MPC Salem Unit 1 Activity Released (Ci) -------------------------------------------------------
: 2)   Will provide for a liquid radwaste discharge rate (as evaluated for each batch release) that is compatible with plant operations (refer to Tables 1-1 and 1-2) .
Nuclide MPC' 1988 -1989 1990 TOTAL CuCi/ml) CURIES CURIES CURIES CURIES -*------------------------------------------Na-24 3E-05 1.38E-02 4.69E-04 1.69E-03 1.60E-02 Cr-51 2E-03 2.38E-02 5.25E-03 1.16£-02 4.06£-02 Mn-54 1E-04 1.01E-01 1.12E-01 1.52E-01 3.65E-01 Fe-59 5E-05 2.66E-04 1.32E-03 1.15E-03 2.73E-03 Co-57 4E-04 4.01E-03 6.11E-03 7.54E-03 1.77E-02 Co-58 9E-05 1.27E+OO 1.82E+OO 1.98E+OO 5.07E+OO Co-60 3E-05 2.77E-01 1. 78E-01 2.39E-01 6.94E-01 Zr-95 6£-05 1.23E-02 1.53E-03 4.52E-03 1.84E-02 Nb-95 1E-04 1.53E-02 3.85E-03 9. 76E-03 2.89£-02 Nb-97 9E-04 2.44E-02 7.94E-05 6.30E-03 3.0SE-02 Tc-99m 3E-03 4.74E-03 4.62E-04 8.53E-04 6.05E-03 Sr-89 3E-06 1.25E-02 1.54E-03 2.38E-03 1.64E-02 Sr-90 3E-07 2.40E-03 6.68E-04 4.66£-04 3.53E-03 Mo-99 4E-05 1.57E-03 N/0 N/D 1.57E-03 Ag-110ni 3E-05 4.96E-03 2.70E-03 8.40E-04 8.50E-03 Sn-113 BE-05 N/D N/D N/D N/D Sb-124 2E-05 6.32E-02 1.36E-02 1.94E-02 9.62E-02 Sb-125 1E-04 9.35E-02 6.53E-02 6.09£-02 2.20E-01 1-131 3E-07 5.54E-02 3.04E-02 3.53E-02 1.21E-01 1-133 1E-06 2.SOE-02 6.88E-03 8.36E-03 4.32E-02 1-134 2E-5 1.10E-02 N/D N/D 1.10E-02
* A-4
* 1-135 4E-06 1.68E-02 1.94E-04 1.42E-04 1. 71E-02 Ce-144 1E-05 1.89E-02 1. 19E-04 1.69E-04 1.92E-02.
* Table A-1 Calculation of Effective MPC Salem Unit 1 Salem ODCM Rev. 8 Activity Released (Ci)
Cs-134 9E-06 1.31E-01 1.16E-01 1.91E-01 4.38E-01 Cs-136 6£-05 9.31E-05 9.79E-04 1.21E-03 2.28E-03 Cs-137 2E-05 1.34E-01 1.28E-01 2.02E-01 4.64E-01 Ba-140 2E-05 2.79£-04 N/D 1.10E-04 3.89E-04 La-140 2E-05 3.89£-04 2.66E-04 5.35E-04 1.19E-03 H-3 3E-03 6.34E+02 6.08E+02 3.53E+02 1.59E+03 Fe-55 SE-04 5.40E-01 1. 75E-01 1.61E-01 8.76E-01 W-187 6£-05 1.25E-02 N/D N/D 1.25E-02 .. 'ii. Zn-65 1E-04 5.49E-04 3.62E-04 7.75E-03 8.66E-03 Zr-97 2E-05 1.37E-02 N/D N/D 1.37E-02 Total C. G1111111a 2.33E+OO 2.49E+OO 2.94E+OO 7.77E+OO Total C. Non-ganma 6.35E+02 6.08E+02 3.53E+02 1.60E+03 MPC, CUCi/111l) 4.71E-06 6.88E-06 9.45E-06 MPC value for 111r .. tricted are* from 10 CFR 20, Appendix B, Table II, CotU111 2. * ** N/D -not detected
Nuclide      MPC' 1988          -1989          1990        TOTAL CuCi/ml)          CURIES        CURIES          CURIES      CURIES Na-24      3E-05          1.38E-02    4.69E-04        1.69E-03    1.60E-02 Cr-51      2E-03          2.38E-02    5.25E-03        1.16£-02      4.06£-02 Mn-54      1E-04          1.01E-01    1.12E-01        1.52E-01      3.65E-01 Fe-59      5E-05          2.66E-04      1.32E-03        1.15E-03      2.73E-03 Co-57      4E-04          4.01E-03      6.11E-03        7.54E-03      1.77E-02 Co-58        9E-05          1.27E+OO      1.82E+OO        1.98E+OO      5.07E+OO Co-60        3E-05          2.77E-01      1. 78E-01      2.39E-01      6.94E-01 Zr-95        6£-05          1.23E-02      1.53E-03        4.52E-03      1.84E-02 Nb-95        1E-04          1.53E-02      3.85E-03        9. 76E-03    2.89£-02 Nb-97        9E-04          2.44E-02      7.94E-05        6.30E-03      3.0SE-02 Tc-99m      3E-03          4.74E-03      4.62E-04        8.53E-04      6.05E-03 Sr-89        3E-06          1.25E-02      1.54E-03        2.38E-03      1.64E-02 Sr-90        3E-07          2.40E-03    6.68E-04        4.66£-04      3.53E-03 Mo-99        4E-05         1.57E-03        N/0            N/D        1.57E-03 Ag-110ni    3E-05          4.96E-03     2.70E-03        8.40E-04      8.50E-03 Sn-113      BE-05              N/D          N/D            N/D            N/D Sb-124      2E-05          6.32E-02      1.36E-02        1.94E-02      9.62E-02 Sb-125      1E-04          9.35E-02    6.53E-02        6.09£-02      2.20E-01 1-131        3E-07          5.54E-02    3.04E-02        3.53E-02      1.21E-01 1-133        1E-06          2.SOE-02    6.88E-03        8.36E-03      4.32E-02 1-134        2E-5          1.10E-02        N/D              N/D      1.10E-02 1-135        4E-06          1.68E-02     1.94E-04        1.42E-04      1. 71E-02 Ce-144      1E-05          1.89E-02     1. 19E-04        1.69E-04      1.92E-02.
* A-5 
Cs-134      9E-06          1.31E-01    1.16E-01        1.91E-01      4.38E-01 Cs-136      6£-05          9.31E-05    9.79E-04         1.21E-03      2.28E-03 Cs-137      2E-05          1.34E-01    1.28E-01        2.02E-01      4.64E-01 Ba-140      2E-05          2.79£-04       N/D          1.10E-04      3.89E-04 La-140      2E-05         3.89£-04    2.66E-04        5.35E-04      1.19E-03 H-3        3E-03         6.34E+02    6.08E+02         3.53E+02     1.59E+03 Fe-55        SE-04         5.40E-01    1. 75E-01        1.61E-01      8.76E-01             ~
* * ** Salem ODCM Table A-2 Calculation of Effective MPC Salem Unit 2 Activity Released (Ci) --------------------------------------------------------
W-187        6£-05          1.25E-02         N/D            N/D        1.25E-02           . 'ii.
Nuclide MPC 1988 1989 1990 TOTAL CuCi/ml) CURIES CURIES CURIES CURIES *-------------------------------------------Na-24 3E-05 1. 04E-02 8.0SE-04 2.28E-03 1.3SE-02 Cr-51 2E-03 3. 1'7E-03 1 .57E-02 1.48E-02 3.37E-02 Mn-54 1E-04 1. 74E-01 1.19E-01 1.52E-01 4.4SE-01 Fe-59 5E-05 2.93E-05 3.00E-03 1 .09E-03 4. 12E-03 Co-57 4E-04 4.55E-03 6.70E-03 7.92E-03 1.92E-02 Co-58 9E-05 1 .32E+OO 2.02E+OO 2.01E+OO S.35E+OO Co-60 3E-05 2.97E-01 2.08E-01 2.36E-01 7.41E-01 . Zr-95 6E-05 3.1SE-03 3.39E-03 5.22E-03 1.18E-02 Nb-95 1E-04 6.55E-03 7.41E-03 1.03E-02 2.42E-02 Nb-97 9E*04 6.92E-03 2.54E-04 5.32E-04 7.71E*03 Tc-99m 3E-03 3.28E-03 6.64E-04 8.66E-04 4.81E*03 Sr-89 3E-06 1.69E*02 1.52E*03 2.28E-03 2.07E-02 Sr-90 3E-07 4.11E-03 6.45E-04 4.73E*04 5.23E-03 Mo-99 4E*OS 1.19E*04 N/D N/D 1.19E-04 Ag*110m 3E*05 1.04E*02 6.41E*03 2.56E-03 1.94E*02 Sn-113 8E*OS N/D N/D N/D N/D Sb-124 2E*OS 5.47E-02 1.89E*02 2.22E*02 9.58E-02 Sb-125 1E*04 9.22E*02 8.0SE-02 7.40E*02 2.47E-01 1*131 3E*07 1.3SE*01 3. 79E-02 3.83E*02 2.11E*01 1-133 1E*06 8.83E*02 8.64E*03 1.07E-02 1.0SE-01 I *134 2E*05 3.49E*02 N/D N/D 3.49£-02 I-135 4E*06 1.90E*02 5.17E-04 7.09E-04 2.02E*02 Ce-144 1E*OS 2.24E*03 6.05E*04 7.67E-05 2.92E*03 Cs-134 9E-06 9.53E*02 1.43E*01 1.86E-01 4.24E*01 Cs-136 6E*OS 2.20E*03 1.39E*03 1.31E*03 4.90E-03 Cs-137 2E*OS 1.09E*01 1.55E*01 1.95E*01 4.59E-01 Ba-140 2E*OS 1.57E*03 N/D N/D 1.57E-03 La-140 2E*OS 1.03E*03 5.19E*04 6.23E*04 2.17E-03 H*3 3E-03 3.68E+02 5.02E+02 3.03E+02 1.17E+03 Fe-55 SE-04 4.69E-01 1.84E*01 2.09E-01 8.62E*01 W*187 6E-05 6.37E-04 N/D N/D 6.37E-04 Zn-65 1E-04 11/D 1.41E*04 1.06E*02 1.07E-02 Total Ci G&1111111 2.48E+OO 2.84E+OO 2.98E+OO 8.30E+OO Total Ci Non-gamna 3.68E+02 5.02E+02 3.03E+02 1.17E+03 MPC 0 (uCi /ml) 3.38E-06 7.85E*06 9.71E-06 MPC value for ...,restricted ere* from 10 CFR 20, Appendix B, Table II, Colllll'I
Zn-65        1E-04          5.49E-04    3.62E-04        7.75E-03     8.66E-03           ~:
: 2. ** N/D -not detected A-6 d Rev. 8 .; ... . t:. ;
Zr-97        2E-05          1.37E-02         N/D           N/D       1.37E-02 Total C. G1111111a        2.33E+OO    2.49E+OO        2.94E+OO      7.77E+OO Total C. Non-ganma        6.35E+02     6.08E+02        3.53E+02     1.60E+03 MPC, CUCi/111l)            4.71E-06      6.88E-06        9.45E-06 MPC value for 111r..tricted are* from 10 CFR 20, Appendix B, Table II, CotU111 2. *
* Salem ODCM Rev. 8 APPENDIX B Technical Basis for Effective Dose Factors Liquid Radioactive Effluent *
  ** N/D - not detected
* A-5
 
d Salem ODCM  Rev.       8 Table A-2 Calculation of Effective MPC Salem Unit 2 Activity Released (Ci)
MPC 1988          1989            1990    TOTAL Nuclide CuCi/ml)        CURIES        CURIES          CURIES    CURIES Na-24 3E-05
: 1. 04E-02 8.0SE-04 2.28E-03 1.3SE-02 Cr-51          2E-03            3. 1'7E-03    1.57E-02        1.48E-02  3.37E-02 Mn-54          1E-04            1. 74E-01      1.19E-01        1.52E-01  4.4SE-01 Fe-59          5E-05            2.93E-05      3.00E-03       1.09E-03  4. 12E-03 Co-57          4E-04           4.55E-03      6.70E-03        7.92E-03  1.92E-02 Co-58          9E-05           1.32E+OO      2.02E+OO        2.01E+OO  S.35E+OO Co-60          3E-05            2.97E-01      2.08E-01        2.36E-01  7.41E-01
  . Zr-95          6E-05           3.1SE-03      3.39E-03        5.22E-03  1.18E-02 Nb-95          1E-04            6.55E-03      7.41E-03        1.03E-02   2.42E-02 Nb-97          9E*04            6.92E-03       2.54E-04        5.32E-04  7.71E*03 Tc-99m        3E-03            3.28E-03      6.64E-04        8.66E-04  4.81E*03 Sr-89          3E-06            1.69E*02      1.52E*03        2.28E-03  2.07E-02 Sr-90          3E-07            4.11E-03      6.45E-04        4.73E*04  5.23E-03 Mo-99          4E*OS            1.19E*04          N/D              N/D    1.19E-04 Ag*110m        3E*05            1.04E*02      6.41E*03        2.56E-03  1.94E*02 Sn-113        8E*OS              N/D          N/D                N/D        N/D Sb-124        2E*OS            5.47E-02      1.89E*02        2.22E*02  9.58E-02 Sb-125        1E*04            9.22E*02      8.0SE-02        7.40E*02  2.47E-01 1*131          3E*07            1.3SE*01      3. 79E-02      3.83E*02  2.11E*01 1-133          1E*06            8.83E*02      8.64E*03        1.07E-02  1.0SE-01 I *134        2E*05            3.49E*02          N/D              N/D  3.49£-02 I-135          4E*06            1.90E*02      5.17E-04        7.09E-04  2.02E*02 Ce-144        1E*OS            2.24E*03      6.05E*04        7.67E-05  2.92E*03 Cs-134        9E-06            9.53E*02      1.43E*01        1.86E-01  4.24E*01 Cs-136        6E*OS            2.20E*03      1.39E*03        1.31E*03  4.90E-03 Cs-137        2E*OS            1.09E*01      1.55E*01        1.95E*01  4.59E-01 Ba-140        2E*OS            1.57E*03          N/D              N/D  1.57E-03 La-140        2E*OS            1.03E*03      5.19E*04        6.23E*04  2.17E-03 H*3          3E-03            3.68E+02      5.02E+02        3.03E+02  1.17E+03 Fe-55          SE-04            4.69E-01      1.84E*01        2.09E-01  8.62E*01 W*187          6E-05            6.37E-04           N/D              N/D  6.37E-04            ....;
Zn-65          1E-04                11/D      1.41E*04        1.06E*02   1.07E-02         . t:.
Total Ci      G&1111111        2.48E+OO      2.84E+OO        2.98E+OO  8.30E+OO Total Ci    Non-gamna          3.68E+02      5.02E+02        3.03E+02  1.17E+03 MPC0 (uCi /ml)                  3.38E-06      7.85E*06        9.71E-06 MPC value for ...,restricted ere* from 10 CFR 20, Appendix B, Table II, Colllll'I 2.
    **    N/D - not detected
**                                                               A-6
* Salem ODCM  Rev. 8 APPENDIX B Technical Basis for Effective Dose Factors Liquid Radioactive Effluent
* B-1
* B-1
* Salem ODCM Rev. 8 APPENDIX B Technical Basis for Effective Dose Factors -Liquid Effluent Releases The radioactive liquid effluents for the years 1982 through 1989 were evaluated to determine the dose contribution of the radionuclide distribution.
* APPENDIX B Salem ODCM Technical Basis for Effective Dose Factors -
This analysis was performed to evaluate the use of a limited dose analysis for determining environmental doses, providing a simplified method of determining compliance with the dose limits of Technical Specification 3.11.1.2.
Liquid Effluent Releases Rev. 8 The radioactive liquid effluents for the years 1982 through 1989 were evaluated to determine the dose contribution of the radionuclide distribution. This analysis was performed to evaluate the use of a limited dose analysis for determining environmental doses, providing a simplified method of determining compliance with the dose limits of Technical Specification 3.11.1.2. For the radionuclide distribution of effluents from Salem, the controlling organ is the GI-LLI. For the last three years the calculated GI-LLI dose is predominately a function of the Fe-55, co-58, Co-60 and
For the radionuclide distribution of effluents from Salem, the controlling organ is the GI-LLI. For the last three years the calculated GI-LLI dose is predominately a function of the Fe-55, co-58, Co-60 and *
* Nb-95 releases. The radionuclides, Co-58 and cs-134 contribute the large majority of the calculated total body dose. The results of the   evaluation for 1989, 1988, and 1987 are presented in Table B-1 and Table B-2.
* Nb-95 releases.
For purposes of simplifying the details of the dose calculational process, it is conservative to identify a controlling, dose significant radionuclide and limit the calculation process to the use of the dose conversion factor for this nuclide. Multiplication of the total release (i.e., cumulative activity for all radi'onuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
The radionuclides, Co-58 and cs-134 contribute the large majority of the calculated total body dose. The results of
For the evaluation of the maximum organ dose, it is conservative to
* the evaluation for 1989, 1988, and 1987 are presented in Table B-1 and Table B-2. For purposes of simplifying the details of the dose calculational process, it is conservative to identify a controlling, dose significant radionuclide and limit the calculation process to the use of the dose conversion factor for this nuclide. Multiplication of the total release (i.e., cumulative activity for all radi'onuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
* B-2
For the evaluation of the maximum organ dose, it is conservative to B-2
* Salem ODCM Rev. 8 use the Nb-95 dose conversion factor (1.51 E+06 mrem/hr per uCi/ml, GI-LLI).     By this approach, the maximum organ dose will be overestimated since this nuclide has the highest organ ddse factor of all the radionuclides evaluated. For the total body calculation, the Fe-59 dose factor (7.27 E+04 mrem/hr per uci/ml, total body) is the highest among the identified dominant nuclides.       For evaluating compliance with the dose limits of Technical Specification 3.11.1.2, the following simplified equations may be used:
* *
Total Body 1.67E-02 '* VOL Dtb  =       cw
* Salem ODCM Rev. 8 use the Nb-95 dose conversion factor (1.51 E+06 mrem/hr per uCi/ml, GI-LLI). By this approach, the maximum organ dose will be overestimated since this nuclide has the highest organ ddse factor of all the radionuclides evaluated.
* A Fe-59,TB *                   (B .1)
For the total body calculation, the Fe-59 dose factor (7.27 E+04 mrem/hr per uci/ml, total body) is the highest among the identified dominant nuclides.
* where:
For evaluating compliance with the dose limits of Technical Specification 3.11.1.2, the following simplified equations may be used: Total Body Dtb = where: Dlb A Fe-59,TB VOL Ci cw 1. 67E-02 1.67E-02 '* VOL = = = = = =
Dlb A  Fe-59,TB
* A Fe-59,TB * (B .1) cw dose to the total body {mrem) 7.27E+04, total body ingestion dose conversion factor for Fe-59 {mrem/hr per uci/ml) ,:\ volume of liquid effluent released {gal}\'* total concentration of all radionuclides (uci/ml) average circulating water discharge rate during release period(gal/min) conversion factor (hr/min) Substituting the value for the Fe-59 total body dose conversion factor, the equation simplifies to: .1.21B+03
                    =
* VOL ----------------
                    =
* (B.2) cw B-3
dose to the total body {mrem) 7.27E+04, total body ingestion dose conversion factor for Fe-59 {mrem/hr per uci/ml)   ,:\
* *
VOL            =  volume of liquid effluent released {gal}\'*
* Salem ODCM Rev. 8 Maximum organ l.67E-02
Ci            =  total concentration of all radionuclides (uci/ml) cw            =  average circulating water discharge rate during release period(gal/min)
* VOL *A Nb-95,GI-LLI Dmu = ---------------------------
: 1. 67E-02      =  conversion factor (hr/min)
* (B.3) where: Dmax A Nb-95,GI-LLI  
Substituting the value for the Fe-59 total body dose conversion factor, the equation simplifies to:
= maximum organ dose (mrem) = 1.51E+06, Gi-LLI ingestion dose conversion factor for Nb-95 (mrem/hr per uCi/ml) Substituting the value for A Nb-95,GI-LLI the equation simplifies to: 2.52E+04
                      .1.21B+03
* VOL Dmax * (B. 4) --cw Tritium is not included in the limited analysis dose assessment for liquid releases, the potential dose resulting normal .. \;. reactor releases is relatively negligible.
* VOL cw      *                         (B.2)
The average annual tritium release from each Salem Unit is approximately 350 curies. The calculated total body dose from such a release is 2.4E-03 mrem/yr via the fish and invertebrate ingestion pathways.
* B-3
This amounts to 0.08% of the design objective dose of 3 mrem/yr. Furthermore, the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste system operation . B-4
* Maximum organ Salem ODCM   Rev. 8 l.67E-02
* Salem ODCM Rev. 8 Table B-1 Adult Dose Contributions Fi sh. and Invertebrate Pathways Unit 1 1989 1988 1987 ................
* VOL *A Nb-95,GI-LLI Dmu = --------------------------- *                     (B.3) where:
----------**----------------------
Dmax             = maximum organ dose (mrem)
----------------------------------
A Nb-95,GI-LLI   = 1.51E+06, Gi-LLI ingestion dose conversion factor for Nb-95 (mrem/hr per uCi/ml)
---------------------------------
Substituting the value for A Nb-95,GI-LLI the equation simplifies to:
Radio* RELEASE TBOOY GI-LLI LIVER RELEASE TBODY GI-LLI LIVER RELEASE TBOOY Gl-LLI LIVER nuclide (Ci) Dose Dose Dose (Ci) Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac.
2.52E+04
Frac ----------------------------------------
* VOL Dmax -                    *                               (B. 4) cw Tritium is not included in the limited analysis dose assessment for liquid releases,   be~ause  the potential dose resulting fro~      normal
--------------------------------................
                                                              .. \;.
MN-54 1.12E-01 0.03 0.06. 0.11 1.01E-01 0.01 0.03 0.03 1.05E-01 a.a1 a.a5 a.a4 FE-55 3.98E-a2 a.as a.a2 a.19 5.44E-01 a.43 a.11 0.76 2.35E-01 a. 16 a.11 a.43 FE-59 1.32E-a3 a.a2 a.a2 a.a3 2.66E-04 * *
reactor releases is relatively negligible. The average annual tritium release from each Salem Unit is approximately 350 curies.
* N/D * *
The calculated total body dose from such a release is 2.4E-03 mrem/yr via the fish and invertebrate ingestion pathways.           This amounts to 0.08% of the design objective dose of 3 mrem/yr.
* C0-58 1.82E+OO a.39 a.56 a.16 1.27E+aO a. 11 a.24 a.a3 1.54E+OO a.17 a.42 a.as C0-60 1. 78E-01 a.11 0.15 a.a4 2..77E-01
Furthermore, the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste system operation .
: a. 10 0.14 0.02 4.21E-01 0. 13 a.31 a.a4 ZN-65 3.62E-a4 a.a1
* B-4
* a.a2 5.49E-a4 a.a1
 
* a.01 N/D * *
Salem ODCM       Rev. 8 Table B-1 Adult Dose Contributions Fi sh. and Invertebrate Pathways Unit 1 1989                                     1988                                   1987 Radio* RELEASE TBOOY                 GI-LLI   LIVER   RELEASE TBODY         GI-LLI LIVER   RELEASE TBOOY           Gl-LLI     LIVER nuclide (Ci)                 Dose     Dose     Dose     (Ci)     Dose       Dose   Dose     (Ci)         Dose     Dose       Dose Frac. Frac. Frac.               Frac. Frac. Frac.                 Frac. Frac.     Frac MN-54             1.12E-01             0.06.
* NB-95 3.8SE-03
0.03              0.11   1.01E-01     0.01     0.03   0.03 1.05E-01       a.a1     a.a5       a.a4 FE-55           3.98E-a2   a.as     a.a2   a.19   5.44E-01     a.43     a.11   0.76 2.35E-01       a. 16     a.11       a.43 FE-59           1.32E-a3   a.a2     a.a2   a.a3   2.66E-04       *         *
* N/D       *           *
* C0-58             1.82E+OO   a.39     a.56   a.16   1.27E+aO     a. 11     a.24   a.a3 1.54E+OO       a.17     a.42       a.as C0-60             1. 78E-01   a.11     0.15   a.a4   2..77E-01     a. 10     0.14   0.02 4.21E-01       0. 13     a.31       a.a4 ZN-65             3.62E-a4   a.a1
* a.a2   5.49E-a4     a.a1
* a.01     N/D         *           *
* NB-95             3.8SE-03
* a. 15
* a. 15
* 1.53E-02
* 1.53E-02
Line 547: Line 816:
* 2.44E-a3
* 2.44E-a3
* a.a8
* a.a8
* AG-110M 2.7aE-a3
* AG-110M           2.7aE-a3
* a.04
* a.04
* 4.96E-03
* 4.96E-03
Line 553: Line 822:
* 2.36E-03
* 2.36E-03
* a.o3
* a.o3
* CS* 134 1.16E-01 0.24
* CS* 134           1.16E-01   0.24
* a.25 1.31E-01 0.17
* a.25   1.31E-01     0.17
* 0.08 3.11E-01 0.34
* 0.08 3.11E-01       0.34
* a.26 CS-137 1.28E-a1 a.16
* a.26 CS-137           1.28E-a1   a.16
* a.21 1.34E-a1 a.10
* a.21   1.34E-a1     a.10
* a.a6 3.a1E-01 a.19
* a.a6 3.a1E-01       a.19
* a.19 Total 2.4aE+aO 2.48E+OO 2.92E+OO Table i-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1989 1988 1987 ** --------------------------------------------
* a.19 Total             2.4aE+aO                             2.48E+OO                             2.92E+OO Table i-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1989                                     1988                                   1987 ISOTOPE           RELEASE   TBOOY     GI-LLI   LIVER     RELEASE TBOOY       Gl*LLI LIVER RELEASE     TBOOY       GI -LLI   LIVER (Ci)     Dose       Dose     Dose       (Ci)     Dose     Dose   Dose     (Ci)         Dose     Dose       Dose Frac.     Frac. Frac.               Frac. Frac. Frac.                 Frac. Frac.     Frac.
----------------------------------
MN-54             1.19E-01 0.02     0.05     o.a9     1.74E-01     0.03     0.07   0.06 1.20E-01 o.a1           a.04       a.a2 FE-55             4.61E-02 0.05         0.02     0.18     4.69E-01     0.42     0.16   0.75 8.74E*01 0.39           0.26       a.72 FE-59             3.00E-03 0.03         0.04     0.06     2.93E-05     *         *
----------------------------------
* N/D         *           *       **
ISOTOPE RELEASE TBOOY GI-LLI LIVER RELEASE TBOOY Gl*LLI LIVER RELEASE TBOOY GI -LLI LIVER (Ci) Dose Dose Dose (Ci) Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac.
C0-58             2.02E+OO 0.37         0.47     a.14     1.32E+OO     0.19     0.29   0.04 1. 71E+OO -0.12         0.31       a.a2 C0-60             2.08E-01 . 0.11       0.13     0.04     2.97E-01     a.12     0.18   0.02 4.23E-01 .. 'i}.a9       0.21       o.a2 ZN-6S             1.41E-04     *
Frac. Frac. Frac. ----------------------------------------
* a.01         N/D       *         *
*---------------------------------------.. -............
* N/D   .~ . *           *
MN-54 1.19E-01 0.02 0.05 o.a9 1.74E-01 0.03 0.07 0.06 1.20E-01 o.a1 a.04 a.a2 FE-55 4.61E-02 0.05 0.02 0.18 4.69E-01 0.42 0.16 0.75 8.74E*01 0.39 0.26 a.72 FE-59 3.00E-03 0.03 0.04 0.06 2.93E-05 * *
* NB-95             7.41E-03
* N/D * * ** C0-58 2.02E+OO 0.37 0.47 a.14 1.32E+OO 0.19 0.29 0.04 1. 71E+OO -0.12 0.31 a.a2 C0-60 2.08E-01 . 0.11 0.13 0.04 2.97E-01 a.12 0.18 0.02 4.23E-01 .. 'i}.a9 0.21 o.a2 ZN-6S 1.41E-04 *
* a.01 N/D * *
* N/D . * *
* NB-95 7.41E-03
* a.22
* a.22
* 6.55E-03
* 6.55E-03
* 0.18
* 0.18
* 7.92E-03
* 7.92E-03
* o. 18
* o. 18 AG-110M           6.41E-03
* AG-110M 6.41E-03
* 0.07
* 0.07
* 1.a4E-02
* 1.a4E-02
* 0.11
* 0.11
* N/D * *
* N/D         *          *        **
* CS-134 1.43E-01 0.25
CS-134           1.43E-01 0.25
* 0.26 9.53E-02 0. 14
* 0.26   9.53E-02     0. 14
* 0.07 3.49E-01 0.25
* 0.07 3.49E-01 0.25
* 0.13 CS-137 1.55E-01 0.16
* 0.13 CS-137           1.55E-01 0.16
* 0.21 1.09E-01.
* 0.21     1.09E-01. 0.09
0.09
* 0.06 3.33E-01 o. 14
* 0.06 3.33E-01 o. 14
* o.a9 Total 2.71E+OO 2.48E+OO 3.82E+OO
* o.a9 Total             2.71E+OO                               2.48E+OO 3.82E+OO
* less than 0.01 N/D = not detected
* less than 0.01 N/D     = not detected
* B-5 Salem OOCM Rev. 8
* B-5
* APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent
 
* c-1
Salem OOCM Rev. 8 APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent
* ** overview Salem OOCM Rev. 8 APPENDIX C Technical Bases for Effective Dose Factors -Gaseous Radioactive Effluents The evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfer factors instead of using dose* factors which are radionuclide specific.
                                              ~.'*
These effective factors, which can be based on typical radionuclide distributions of releases, can be applied to the total radioactivity released to approximate the dose in the environment (i.e., instead of having to perform individual radionuclide dose analyses only a single multiplication (Kcffi Meff or Neff) times the total quantity of radioactive material released would be needed)
* c-1
* This approach provides a reasonable estimate of the actual dose while eliminating the need for a detailed calculational technique. : Determination of Effective Dose Factors Effective dose transfer factors are calculated by the following equations:  
 
--where: * (C.1) = the effective total body dose factor due to gamma emissions from all noble gases released = the total body dose factor due to gamma emissions from each noble gas radionuclide i released = the fractional abundance of noble gas radionuclide i relative to the total noble gas activity c-2
Salem OOCM Rev. 8
* where: where: where: Salem OOCM Rev. 8 (C. 2) (L + 1.1 M)eff =*the effective skin dose factor due to beta and gamma emissions from all noble gases released = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released + 1.1 MctJ (C. 3) = the effective air dose factor due to gamma emissions from all noble gases released = the air dose factor due to gamma emissions from each noble gas radionuclide i released (C. 4) = the effective air dose factor due to beta emissions from all noble gases released = the air dose factor due to beta each noble gas radionuclide i released Normally, it would be expected that past radioactive effluent data would be used for the determination of the effective dose factors. However, the noble gas releases from Salem have been maintained to such negligible quantities that the inherent variability in the data makes any meaningful evaluations difficult.
* overview APPENDIX C Technical Bases for Effective Dose Factors -
For the past years, the total noble* gas releases have been limited to* 2,000 Ci for 1984, C-3
Gaseous Radioactive Effluents The evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfer factors instead of using dose* factors which are radionuclide specific. These effective factors, which can be based on typical radionuclide distributions of releases, can be applied to the total radioactivity released to approximate the dose in the environment (i.e., instead of having to perform individual radionuclide dose analyses only a single multiplication (Kcffi Meff or Neff) times the total quantity of radioactive material released would be needed)
*
* This approach provides a reasonable estimate of the actual dose while eliminating the need for a detailed calculational technique.
* Salem ODCM Rev. 8 2,800 Ci for 1985, 2,700 Ci for 1986, 1700 Ci for 1988, and 1500 Ci for 1989. Therefore, in order to provids a reasonable basis for the derivation of the effective noble gas dose factors, the primary coolant source term from ANSI N237-1976/ANS-18.1, "Source Term Specifications," has been used as representing a typical distribution.
                                                              ~ :
The effective dose factors as derived are presented in Table C-1. Application To provide an additional degree of conservatism, a factor of 0.50 is introduced into the dose calculational process when the effective
Determination of Effective Dose Factors Effective dose transfer factors are calculated by the following equations:
* dose transfer factor is used. This provides additional assurance that the evaluation of doses by the use of a single effective factor will not significantly underestimate any actual doses in the environment.
              -         *                             (C.1) where:
For evaluating compliance with:the dose limits of Technical Specification 3.11.2.2, the following simplified equations may be used: 3.17B-08 Di: = --------* X/Q
                    = the effective total body dose factor due to gamma emissions from all noble gases released
* M.:tr * :E Qi (C. 5) o.so and 3.17E-08 = --------* X/Q
                  =   the total body dose factor due to gamma emissions from each noble gas radionuclide i released
* Neff * :E Qi (C. 6) o.so C-4 
                    = the fractional abundance of noble gas radionuclide i relative to the total noble gas activity
* . * ** .... where: Dg Db X/Q Melf Neff Qi 3.17E-08 0.50 = = = = = = = Salem ODCM Rev. 8 air dose due to gamma emissions for the cumulative release of all noble gases (mrad) air dose due to beta emissions for the cumulative releas.e of all noble gases (mrad) atmospheric dispersion to the controlling site boundary (sec/m3) 5.3E+02, effective gamma-air dose factor (mrad/yr per uCi/m3) 1.1E+03, effective beta-air dose factor (mrad/yr per uCi/m3) cumulative release for all noble gas radionuclides (uCi} conversion factor (yr/sec} conservatism factor to account for the variability in the effluent data Combining the the dose calculational equations simplify to: DI = 3.SE-05
**                                    c-2
 
Salem OOCM   Rev. 8
*                                                                (C. 2) where:
(L + 1.1 M)eff   =*the effective skin dose factor due to beta and gamma emissions from all noble gases released
(~  +  1.1 ~)    = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released (C. 3) where:
MctJ  =   the effective air dose factor due to gamma emissions from all noble gases released
                =   the air dose factor due to gamma emissions from each noble gas radionuclide i released (C. 4) where:
                =   the effective air dose factor due to beta emissions from all noble gases released
                =   the air dose factor due to beta emissi~ns f~pm each noble gas radionuclide i released Normally, it would be expected that past radioactive effluent data would be used for the determination of the effective dose factors.
However, the noble gas releases from Salem have been maintained to such negligible quantities that the inherent variability in the data makes any meaningful evaluations difficult.       For the past years, the total noble* gas releases have been limited to* 2,000 Ci for 1984, C-3
 
Salem ODCM Rev. 8
* 2,800 Ci for 1985, 2,700 Ci for 1986, 1700 Ci for 1988, and 1500 Ci for 1989. Therefore, in order to provids a reasonable basis for the derivation of the effective noble gas dose factors, the primary coolant source term from ANSI N237-1976/ANS-18.1, "Source Term Specifications," has been used as representing a typical distribution. The effective dose factors as derived are presented in Table C-1.
Application To provide an additional degree of conservatism, a factor of 0.50 is introduced into the dose calculational process when the effective dose transfer factor is used. This conservat~sm  provides additional assurance that the evaluation of doses by the use of a single effective factor will not significantly underestimate any actual doses in the environment. For evaluating compliance with:the dose limits of Technical Specification 3.11.2.2, the following simplified equations may be used:
3.17B-08 Di:   = --------
o.so
* X/Q
* X/Q
* E Qi (C. 7) and = 7.0E-05
* M.:tr *  :E Qi             (C. 5) and 3.17E-08
  ~      = --------
o.so
* X/Q
* X/Q
* E Qi
* Neff
: 8) The effective dose factors are used on a very limited basis for the purpose of facilitating the timely assessment of radioactive effluent releases, particularly during periods of computer malfunction where a detailed dose assessment may be unavailable . c-s
                                        *  :E Qi             (C. 6)
* Salem ODCM Rev. 3 Table C-1 Effective Dose Factors Noble Gases -Total Body and Skin Radionuclide Kr-85 Kr-88 Xe-133m Xe-133 Xe-135 Total Noble Gases -Air Radionuclide Kr-85 Kr-88 Xe-133m Xe-133 Xe-135 Total 0.01 0.01 0.01 0.95 0.02 0.01 0.01 0.01 0.95 0.02 Total Body Effective Dose Factor K.tr (mrem/yr per uCi/m 3) l.5E+02 2.5E+OO 3.0E+02 3.6E+Ol 4.8E+02 Gamma Air Effective Dose Factor M.tr (mrad/yr per l.5E+02 3.3E+OO 3 . .4E+02 3.8E+Ol 5.3E+02 SkinEffective Dose Factor ( L+ 1. 1 M) etr (mrem/yr per uCi/m 3) l.4E+Ol l.9E+02 l.4E+Ol 6.6E+02 7.9E+Ol 9.6E+02 Beta Air Effective Dose Factor Neff (mrad/yr per -uci/m 3) 2.0E+Ol 2.9E+Ol 1. SE+Ol 1. OE+03 4 :;9E+Ol 1. 1E+03 Term
* C-4
* Based on Noble gas distribution from ANSI N237-1976/ANSI-18.l, "Source Specifications." C-6
 
* ., Salem ODCM Rev. 8 APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1
Salem ODCM  Rev. 8
* *
* where:
* Salem ODCM Rev. 8 APPENDIX D Technical Basis for* Effective Dose Parameter Gaseous Radioactive Effluent Releases The pathway dose factors for the controlling infant age group were evaluated to determine the controlling pathway, organ and radionuclide.
Dg        =    air dose due to gamma emissions for the cumulative release of all noble gases (mrad)
This analysis was performed to provide a simplified method for determining compliance with Technical Specification 3.11.2.3 For the infant age group, the controlling pathway is the grass-milk-cow (g/m/c) pathway. An infant receives a greater radiation dose from the g/m/c pathway than any other pathway. Of this g/m/c pathway, the maximum exposed organ including the total body, is the thyroid, and the highest dose contributor is radionuclide I-131. The results for this evaluation are presented in Table D-1. For purposes of simplifying the details of the dose calculation process, it is conservative to identify a controlling, dose significant organ and radionuclide and limit the calculation process to the use of the dose conversion factor for the organ and radionuclide.
Db            air dose due to beta emissions for the cumulative releas.e of all noble gases (mrad)
Multiplication of the total release (i.e. cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative . D-2
X/Q      =    atmospheric dispersion to the controlling site boundary (sec/m3)
* * ., Salem ODCM Rev. 3 For the evaluation of the dose commitment via a controlling pathway and age group, it is conservative to use the infant, g/m/c, thyroid, I-131 pathway dose factor (1.05E12 m 2 mrem/yr per uCi/sec).
Melf      =    5.3E+02, effective gamma-air dose factor (mrad/yr per uCi/m3)
By this approach, the maximum dose commitment will be overestimated since I-131 has the highest pathway dose factor of all radionuclides evaluated.
Neff      =    1.1E+03, effective beta-air dose factor (mrad/yr per uCi/m3)
For evaluating compliance with the dose limits of Technical Specification 3.11.2.3, the following simplified equation may be used: where: Dmax w X/Q D/Q Qi 3.17E-8 RI-131 Dmu: = 3.17E-8
Qi        =    cumulative release for all noble gas radionuclides (uCi}
3.17E-08 =    conversion factor (yr/sec}
0.50      =    conservatism factor to account for the variability in the effluent data Combining the  constant~,  the dose calculational equations simplify to:
DI  =  3.SE-05 and
* X/Q
* E Qi                  (C. 7)
                  =  7.0E-05
* X/Q
* E Qi                *~JC. 8)
The effective dose factors are used on a very limited basis for the purpose of facilitating the timely assessment of radioactive effluent releases, particularly during periods of computer malfunction where a detailed dose assessment may be unavailable .
c-s
 
Salem ODCM       Rev. 3
* Table C-1 Effective Dose Factors Noble Gases - Total Body and Skin Total Body Effective       SkinEffective Radionuclide                          Dose Factor              Dose Factor K.tr                 ( L+ 1. 1 M) etr (mrem/yr per uCi/m3 )    (mrem/yr per uCi/m3 )
Kr-85              0.01                                              l.4E+Ol Kr-88              0.01                  l.5E+02                     l.9E+02 Xe-133m            0.01                  2.5E+OO                     l.4E+Ol Xe-133              0.95                  3.0E+02                      6.6E+02 Xe-135              0.02                  3.6E+Ol                     7.9E+Ol Total                                    4.8E+02                     9.6E+02 Noble Gases - Air Gamma Air Effective    Beta Air Effective Radionuclide                          Dose Factor          Dose Factor M.tr                    Neff (mrad/yr per uCi/~)    (mrad/yr per -uci/m3 )
Kr-85              0.01                                            2.0E+Ol Kr-88              0.01                  l.5E+02                    2.9E+Ol Xe-133m            0.01                  3.3E+OO                    1. SE+Ol Xe-133              0.95                  3 . .4E+02                1. OE+03 Xe-135              0.02                  3.8E+Ol                    4 :;9E+Ol Total                                    5.3E+02                    1. 1E+03
* Based on Noble gas distribution from ANSI N237-1976/ANSI-18.l,       "Source Term    Specifications."
C-6
 
Salem ODCM Rev. 8 APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1
 
Salem ODCM Rev. 8
* APPENDIX D Technical Basis for* Effective Dose Parameter Gaseous Radioactive Effluent Releases The pathway dose factors for the controlling infant age group were evaluated to determine the controlling pathway, organ and radionuclide. This analysis was performed to provide a simplified method for determining compliance with Technical Specification 3.11.2.3   For the infant age group, the controlling pathway is the grass-milk-cow (g/m/c) pathway. An infant receives a greater radiation dose from the g/m/c pathway than any other pathway. Of this g/m/c pathway, the maximum exposed organ including the total
* body, is the thyroid, and the highest dose contributor is radionuclide I-131.
in Table D-1.
The results for this evaluation are presented For purposes of simplifying the details of the dose calculation process, it is conservative to identify a controlling, dose significant organ and radionuclide and limit the calculation process to the use of the dose conversion factor for the organ and radionuclide. Multiplication of the total release (i.e. cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative .
* D-2
 
Salem ODCM   Rev. 3
* For the evaluation of the dose commitment via a controlling pathway and age group, it is conservative to use the infant, g/m/c, thyroid, I-131 pathway dose factor (1.05E12 m2 mrem/yr per uCi/sec).     By this approach, the maximum dose commitment will be overestimated since I-131 has the highest pathway dose factor of all radionuclides evaluated.
For evaluating compliance with the dose limits of Technical Specification 3.11.2.3, the following simplified equation may be used:
* where:
Dmax w
Dmu:
                    =
                    =
                          = 3.17E-8
* W
* W
* RI-131
* RI-131 maximum organ.dose (mrem)
* E Qi = = = = = = = = maximum organ.dose (mrem) atmospheric dispersion parameters to the *controlling location(s) as Table 3.2-4. , atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways (sec/m 3) atmospheric deposition for vegetation, milk and ground plane exposure pathways (m.2) cumulative release over the period of interest for radioiodines and particulates conversion factor (yr/sec) I-131 dose parameter for the thyroid for the identified controlling pathway 1.05E12 (m 2 mrem/yr per uCi/sec), infant thyroid dose parameter with the cow-milk=grass pathway controlling The ground plane exposure and inhalation pathways need not be considered when the above simplified calculation method is used because for the overall negligible contribution of these pathways to D-3
* E Qi atmospheric dispersion parameters to the
* *
                        *controlling location(s) as identified~~n Table 3.2-4.                                 ,
* Salem ODCM Rev. 3 the total thyroid dose. It is recognized that for some particulate radioiodines (e.g., Co-60 and Cs-137), the ground exposure pathway may represent a higher dose contribution than either the vegetation or milk pathway. However, use of the I-131 thyroid dose parameter for all radionuclides will the organ dose calculation, especially considering that no other radionuclide has a higher dose parameter for any organ via any pathway than I-131 for the thyroid via the milk pathway (see Table D-1)
X/Q  =  atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways (sec/m3 )
* The location of pathways and the maximum organ so calculation may be based on the available pathways in the surrounding environment of Salem as identified by the annual land-use census (Technical Specification.3.12.2).
D/Q  =    atmospheric deposition for vegetation, milk and ground plane exposure pathways (m.2 )
Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4
Qi  =  cumulative release over the period of interest for radioiodines and particulates 3.17E-8  =    conversion factor (yr/sec)
* D-4
RI-131  =    I-131 dose parameter for the thyroid for the identified controlling pathway
* *
                    =    1.05E12 (m2 mrem/yr per uCi/sec), infant thyroid dose parameter with the cow-milk=grass pathway controlling The ground plane exposure and inhalation pathways need not be
* Target Organs Total Body Liver Thyroid Kidney Lung GI-LLI Salem OOCM Rev. 8 Table D-1 Infant Dose Contributions Fraction of Total Organ and Body Dose PATHWAYS Grass-Cow-Milk 0.02 0.23 0.59 0.02 0.01 0.02 Ground Plane 0.15 0.14 0.15 0.15 0.02 0.15 Fraction of Dose Contribution .Q:l Pathway Pathway Grass-Cow-Milk 0.92 Ground Plane 0.08 Inhalation
., considered when the above simplified calculation method is used because for the overall negligible contribution of these pathways to D-3
* D-5 Salem ODCM Rev. 8
 
* APPENDIX E Radiological Environmental Monitoring Program Sample Type, Location and Analysis * **. \:
Salem ODCM Rev. 3
* E-1
* the total thyroid dose. It is recognized that for some particulate radioiodines (e.g., Co-60 and Cs-137), the ground exposure pathway may represent a higher dose contribution than either the vegetation or milk pathway. However, use of the I-131 thyroid dose parameter for all radionuclides will ma~imize the organ dose calculation, especially considering that no other radionuclide has a higher dose parameter for any organ via any pathway than I-131 for the thyroid via the milk pathway (see Table D-1)
* *
* The location of exposur~ pathways and the maximum organ so calculation may be based on the available pathways in the
* Salem ODCM Rev. 3 APPENDIX E SAMPLE DESIGNATION Samples are identified by a three part code. The first two letters are the power station identification code, in this case "SA". The next three letters are for the media sampled. AIO = Air Iodine IDM = Immersion Dose (TLD) APT = Air Particulates MLK = Milk ECH = Hard Shell Blue Crab PWR = Potable Water (Raw) ESF = Edible Fish PWT = Potable Water (Treated)
* surrounding environment of Salem as identified by the annual land-use census (Technical Specification.3.12.2). Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4 *
ESS = Sediment RWA = Rain Water (Precipitation)
* D-4
FPB = Beef SWA = Surf ace Water FPL = Green Leafy Vegetables VGT = Fodder Crops (Various)
 
FPV = Vegetable (Various)
Salem OOCM   Rev. 8
WWA = Well Water GAM = Game The last four symbols are a location code based on direction and distance from the site. Of these, the first two represent each of the sixteen angular sectors of 22.5 degrees centered about the reactor site. Sector one is divided evenly by the north axis and other sectors are numbered in a clockwise direction; i.e., 2=NNE, 3=NE, 4=ENG, etc. The next digit is a letter which represents the radical distance from the plant: s = on.:..site location E = 4-5 miles off-site A = 0-1 miles off-site F = 5-10 miles off-site B = 1-2 miles off-site G = 10-20 miles c = 2-3 miles off-site H = > 20 miles off-sit. D = 3-4 miles off-site The last number is the station numerical designation within each sector and zone; e.g., 1,2,3, ... For example; the designation SA-WWA-501 would indicate a sample in the SGS program (SA), consisting of well water (WWA), which had been collected in sector number 5, centered at 90' (due east) with respect to the.reactor site at a radical distance of 3 to 4 miles off-site; (therefore, radial distance D). The number 1 indicated that this is sampling station #1 in that particular sector . E-2
* Table D-1 Infant Dose Contributions Fraction of Total Organ and Body Dose PATHWAYS Target Organs          Grass-Cow-Milk                 Ground Plane Total Body                  0.02                         0.15 Liver                      0.23                          0.14 Thyroid                    0.59                          0.15 Kidney                      0.02                         0.15 Lung                        0.01                          0.02 GI-LLI                      0.02                         0.15 Fraction of Dose Contribution .Q:l Pathway Pathway Grass-Cow-Milk         0.92 Ground Plane           0.08 Inhalation               *
* *
* D-5
* Salem ODCM Rev. 3 SAMPLING LOCATIONS All sampling locations and specific information about the individual locations are given in Table E. Maps E-1 and E-2 show the locations of sampling stations with respect to the site. TABLE E-1 STATION CODE STATION LOCATION 2S2 0.4 mi. NNE of vent 3S3 700 ft. NNE of vent; fresh water holding tank 5Sl 1.0 mi. E of vent; site access road 6S2 0.2 mi. ESE of vent; observation building 7SI lOSl llSl llAl 15Al 16Al 12Cl. 4D2 5Dl lODl 0.12 mi. SE of vent; station personnel gate 0.14 mi. SSW of vent; site shoreline.
 
0.09 mi. SW of vent; site shoreline 0.2 mi. W of vent; outfall area 0.3 mi. NW of vent; cooling tower blowdown discharge line 0.7 mi. NNW of vent; south storm drain discharge line 2.5 mi. WSW of vent; west bank of Delaware River 3.7 mi. ENE of vent; Alloway Creek Neck Road 3.5 mi. E of vent; local farm 3.9 mi. SSW of vent; Taylor's Bridge Spur E-3 SAMPLE TYPES IDM WWA AIO, APT, IDM IDM IDM IDM IDM ECH, ESF, ESS, SWA ** 'te. ESS ESS ECH, ESF, ESS SWA IDM AIO, APT, IDM, WWA IDM Salem ODCM 3
Salem ODCM Rev. 8 APPENDIX E Radiological Environmental Monitoring Program Sample Type, Location and Analysis
* TABLE E-1 (Cont'd) STATION CODE STATION LOCATION SAMPLE TYPES llDl 3.5 mi. SW of vent GAM 14Dl 3.4 mi. WNW of vent; Bay View, Delaware IDM 2El 4.4 mi. NNE of vent; local farm IDM 3El 4.1 mi. NE of vent; local FPB, FPV, GAM, IDM, VGT, WWA 3F2 5.7 mi. NE of vent; local farm FPV 7El 4.5 mi. SE of vent; 1 mi. W of Mad. ESF, ESS, SWA Horse Creek 9El 5.0 mi. SW of vent IDM
                                                  \:
* 11E2 5.0 mi. SW of vent IDM 12El 4.4 mi. WSW of vent; Thomas Landing IDM 13El 4.2 mi. w of vent; Diehl House Lab IDM 13E3 4.9 mi. w of vent; local VGT '* 1* ,* 16El 4.1 mi. NNW of vent; Port Penn AIO, APT, IDM lFl 5.8 mi. N of vent; Fort Elf sborg AIO, APT, IDM 1F2 7.1 mi. N of vent; midpoint of SWA Delaware 1F3 5. 9' mi. N of vent; local farm FPL, FPV 2F2 8.7:mi. NNE of vent; Salem Substation AIO, APT, IDM, RWA 2F3 8.0 mi. NNE of vent; local farm FPV 2F4 6.3 mi. NNE of vent; local FPV 2F5 7.5 mi. NNE of vent; Salem High School IDM ** E-4
* E-1
* *
 
* STATION CODE 2F6 2F7 3F2 3F3 5Fl 5F2 6Fl 10F2 llFl 11F3 12Fl 13F2 13F3 13F4 14Fl 14F2 14F3 14F4 15F3 Salem ODCM Rev. 8 TABLE E-1 (Cont'd) STATION LOCATION 7.3 mi. NNE of vent; Southern Training center 5.7 mi. NNE of vent; local farm 5.1 mi. NE of vent; Hancocks Bridge Municipal Building 8.6 mi. NE of vent; Quinton Township School 6.5 mi. E of vent 7.0 mi. E of vent; local farm 6.4 mi. ESE of vent; Stow Neck Road 9.1 mi. SE of vent; Bayside, NJ 5.8 mi. SSW of vent 6.2 mi. SW of vent; Taylor's Bridge Delaware 5.3 mi. SW of vent; Townsend, DE 9.4 mi. WSW of vent; Townsend Elem. School 6.5 mi. W of vent; Odessa, DE 9.3 mi. W of vent; Redding Middle School, Middletown, DE 9.8 mi. W of vent; Middletown, DE 5.5 mi. WNW of vent; local farm 6.6 mi. WNW of vent; Boyds Corner 5.4 mi. WNW of vent; local farm 7.6 mi. WNW of vent; local farm 5.4 mi. NW of vent E-5 SAMPLE TYPES IDM MLK, VGT IDM IDM FPV,IDM VGT IDM IDM IDM IDM MLK, VGT IDM' IDM IDM VGT IDM FPV MLK IDM I I i *I 
Salem ODCM Rev. 3
'-* * .Salem ODCM Rev. 8 STATION CODE 16Fl 16F2 lGl 1G3 2Gl TABLB B-1 (Cont'd) STATION LOCATION 6.9 mi. NNW of vent; C&D Canal 8.1 mi. NNW of vent; Delaware City Public School 10.3 mi. N of vent; local farm 19 mi. N of vent; Wilmington, DE 12 mi. NNE of vent; Mannington Township, NJ 3Gl 17 mi. NE of vent; local farm lOGl 12 mi. SSW of vent; Smyrna, DE 16Gl 15 mi. NNW of vent; Greater Wilminqton Airport 3Hl 3H3 3H5 . 32 mi. NE of vent; National Park, NJ 110 mi. NE of vent; Research and Testing 25 mi. NE of vent; local farm SAMPLE TYPES ESS, SWA IDM FPV IDM FPV IDM, MLK, VGT IDM IDM IDM AIO, APT, IDM FPL,. FPV 
* APPENDIX E SAMPLE DESIGNATION Samples are identified by a three part code. The first two letters are the power station identification code, in this case "SA". The next three letters are for the media sampled.
* * -* Salem_ OOCM Rev. 8 SAMPLES COLLECTION AND ANALYSIS Sample Collection Method Air Particulate Continuous low volume air sampler. Sample collected every week along with the filter change. Air Iodine Crab and Fish Sediment Direct A TEDA impregnated charcoal cartridge is connected to air particulated air sampler and is collected weekly *at filter change. Two batch samples are sealed in a plastic bag or jar and frozen semi-annually or when in season. A sediment sample is taken semi-annually.
AIO = Air Iodine                   IDM   = Immersion Dose (TLD)
2 TLD's will be. collected from each location quarterly.
APT = Air Particulates             MLK   = Milk ECH = Hard Shell Blue Crab         PWR   = Potable Water (Raw)
E-7 Analysis Gross Beta analysis on each weekly sample. Gamma spectrometry shall be performed if gross beta exceeds 10 times the yearly mean of the control station value. As well one sample is analyzed > 24 hrs after sampling to allow for radon and thorium daughter decay. Gamma isotopic analysis on quarterly composites.
ESF = Edible Fish                 PWT   = Potable Water (Treated)
Iodine 131 analysis are performed on each weekly sample. .;. Gamma *isotopj_..Ct:
ESS = Sediment                     RWA   = Rain Water (Precipitation)
analysis of edible portion on collection.
FPB =   Beef                         SWA   = Surf ace Water FPL =   Green Leafy Vegetables       VGT   = Fodder Crops (Various)
Gamma isotopic analysis semi-annually.
FPV =   Vegetable (Various)         WWA   = Well Water GAM =   Game The last four symbols are a location code based on direction and distance from the site. Of these, the first two represent each of the sixteen angular sectors of 22.5 degrees centered about the reactor site. Sector one is divided evenly by the north axis and
Gamma dose quarterly.
* other sectors are numbered in a clockwise direction; i.e., 2=NNE, 3=NE, 4=ENG, etc. The next digit is a letter which represents the radical distance from the plant:
* * ** Sample Milk Water (Rain, Potable, surf ace) Salem ODCM Rev. 8 SAMPLE COLLECTION AND ANALYSIS (Cont'd) Collection Method Sample of fresh milk is collected for each farm semi-monthly when cows are in pasture,
s =
* monthly at other times. Sample to be collected monthly providinq winter icinq conditions allow. .. E-8 Analysis Gamma isotopic analysis and I-131 analysis on each sample on collection.
A =
Gamma isotopic monthly H-3 on quarterly surface sample, monthly on qround water sample . . ... .}}
on.:..site location 0-1 miles off-site E
F
                                          =
                                          =
4-5 miles off-site 5-10 miles off-site B = 1-2 miles off-site               G =   10-20 miles off-si~e c = 2-3 miles off-site               H =   > 20 miles off-sit.
D = 3-4 miles off-site The last number is the station numerical designation within each sector and zone; e.g., 1,2,3, ... For example; the designation SA-WWA-501 would indicate a sample in the SGS program (SA),
consisting of well water (WWA), which had been collected in sector number 5, centered at 90' (due east) with respect to the.reactor site at a radical distance of 3 to 4 miles off-site; (therefore, radial distance D). The number 1 indicated that this is sampling station #1 in that particular sector .
* E-2
 
Salem ODCM   Rev. 3
* SAMPLING LOCATIONS All sampling locations and specific information about the individual locations are given in Table E. Maps E-1 and E-2 show the locations of sampling stations with respect to the site.
TABLE E-1 STATION CODE       STATION LOCATION                   SAMPLE TYPES 2S2     0.4 mi. NNE of vent                           IDM 3S3     700 ft. NNE of vent; fresh water             WWA holding tank 5Sl     1.0 mi. E of vent; site access road         AIO, APT, IDM 6S2     0.2 mi. ESE of vent; observation             IDM building 7SI       0.12 mi. SE of vent; station personnel       IDM gate
* lOSl llSl llAl 0.14 mi. SSW of vent; site shoreline.
0.09 mi. SW of vent; site shoreline 0.2 mi. W of vent; outfall area IDM IDM ECH, ESF, ESS, SWA **
                                                              'te.
15Al      0.3 mi. NW of vent; cooling tower           ESS blowdown discharge line 16Al      0.7 mi. NNW of vent; south storm drain       ESS discharge line 12Cl. 2.5 mi. WSW of vent; west bank of           ECH, ESF, ESS Delaware River                               SWA 4D2      3.7 mi. ENE of vent; Alloway Creek           IDM Neck Road 5Dl      3.5 mi. E of vent; local farm               AIO, APT, IDM, WWA lODl      3.9 mi. SSW of vent; Taylor's Bridge         IDM Spur
* E-3
 
Salem ODCM  ~ev. 3
* TABLE E-1 (Cont'd)
STATION CODE        STATION LOCATION                      SAMPLE TYPES llDl    3.5 mi. SW of vent                              GAM 14Dl    3.4 mi. WNW of vent; Bay View, Delaware        IDM 2El      4.4 mi. NNE of vent; local farm                IDM 3El      4.1 mi. NE of vent; local                      FPB, FPV, GAM, IDM, VGT, WWA 3F2      5.7 mi. NE of vent; local farm                  FPV 7El      4.5 mi. SE of vent; 1 mi. W of Mad.            ESF, ESS, SWA Horse Creek 9El      5.0 mi. SW of vent                             IDM 11E2    5.0 mi. SW of vent                             IDM 12El    4.4 mi. WSW of vent; Thomas Landing            IDM 13El    4.2 mi. w of vent; Diehl House Lab            IDM 13E3    4.9 mi. w of vent; local                      VGT  '*
1*
16El    4.1 mi. NNW of vent; Port Penn                  AIO, APT, IDM lFl      5.8 mi. N of vent; Fort Elf sborg              AIO, APT, IDM 1F2      7.1 mi. N of vent; midpoint of                  SWA Delaware 1F3      5. 9' mi. N of vent; local farm                FPL, FPV 2F2      8.7:mi. NNE of vent; Salem Substation          AIO, APT, IDM, RWA 2F3      8.0 mi. NNE of vent; local farm                FPV 2F4      6.3 mi. NNE of vent; local                      FPV 2F5      7.5 mi. NNE of vent; Salem High School          IDM
**                                    E-4
 
Salem ODCM  Rev. 8 TABLE E-1 (Cont'd)
* STATION CODE 2F6 STATION LOCATION 7.3 mi. NNE of vent; Southern Training center SAMPLE TYPES IDM 2F7    5.7 mi. NNE of vent; local farm                 MLK, VGT 3F2    5.1 mi. NE of vent; Hancocks Bridge              IDM Municipal Building 3F3    8.6 mi. NE of vent; Quinton Township            IDM School 5Fl      6.5 mi. E of vent                               FPV,IDM 5F2      7.0 mi. E of vent; local farm                  VGT 6Fl      6.4 mi. ESE of vent; Stow Neck Road            IDM 7F~      9.1 mi. SE of vent; Bayside, NJ                IDM 10F2    5.8 mi. SSW of vent                             IDM llFl    6.2 mi. SW of vent; Taylor's Bridge            IDM Delaware 11F3    5.3 mi. SW of vent; Townsend, DE                MLK, VGT 12Fl    9.4 mi. WSW of vent; Townsend Elem.            IDM' School 13F2    6.5 mi. W of vent; Odessa, DE 13F3    9.3 mi. W of vent; Redding Middle              IDM School, Middletown, DE 13F4    9.8 mi. W of vent; Middletown, DE              IDM 14Fl    5.5 mi. WNW of vent; local farm                VGT 14F2    6.6 mi. WNW of vent; Boyds Corner              IDM 14F3    5.4 mi. WNW of vent; local farm                FPV 14F4    7.6 mi. WNW of vent; local farm                MLK 15F3    5.4 mi. NW of vent                             IDM
* E-5                              *I I
I i
 
                                              .Salem ODCM  Rev. 8
'-* STATION CODE TABLB B-1 (Cont'd)
STATION LOCATION            SAMPLE TYPES 16Fl    6.9 mi. NNW of vent; C&D Canal              ESS, SWA 16F2    8.1 mi. NNW of vent; Delaware City          IDM Public School lGl    10.3 mi. N of vent; local farm               FPV 1G3    19 mi. N of vent; Wilmington, DE            IDM 2Gl    12 mi. NNE of vent; Mannington              FPV Township, NJ 3Gl    17 mi. NE of vent; local farm               IDM, MLK, VGT lOGl    12 mi. SSW of vent; Smyrna, DE              IDM 16Gl    15 mi. NNW of vent; Greater Wilminqton      IDM Airport
* 3Hl 3H3 3H5 32 mi. NE of vent; National Park, NJ 110 mi. NE of vent; Research and Testing 25 mi. NE of vent; local farm IDM AIO, APT, IDM FPL,. FPV
 
Salem_ OOCM    Rev. 8
* Sample SAMPLES COLLECTION AND ANALYSIS Collection Method            Analysis Air Particulate  Continuous low volume      Gross Beta analysis air sampler. Sample        on each weekly collected every week      sample. Gamma along with the filter      spectrometry shall change.                     be performed if gross beta exceeds 10 times the yearly mean of the control station value. As well one sample is analyzed > 24 hrs after sampling to allow for radon and thorium daughter decay. Gamma isotopic analysis on quarterly composites.
Air Iodine      A TEDA impregnated          Iodine 131 analysis charcoal cartridge is      are performed on connected to air            each weekly sample.
particulated air sampler and is collected weekly
                  *at filter change.
Crab and Fish    Two batch samples are      Gamma *isotopj_..Ct:
sealed in a plastic         analysis of edible bag or jar and frozen       portion on collection.
semi-annually or when in season.
Sediment        A sediment sample is         Gamma isotopic taken semi-annually.        analysis semi-annually.
Direct          2 TLD's will be.           Gamma dose quarterly.
collected from each location quarterly.
-*                                  E-7
 
Salem ODCM Rev. 8
* Sample SAMPLE COLLECTION AND ANALYSIS (Cont'd)
Collection Method          Analysis Milk          Sample of fresh milk    Gamma isotopic is collected for each    analysis and I-131 farm semi-monthly when  analysis on each cows are in pasture,
* sample on collection.
monthly at other times.
Water (Rain,  Sample to be collected  Gamma isotopic Potable,      monthly providinq winter monthly H-3 on surf ace)      icinq conditions allow. quarterly surface sample, monthly on qround water sample .
                                                        ....~.
**                               E-8}}

Latest revision as of 18:06, 28 February 2020

Rev 8 to Salem Nuclear Generating Station Odcm.
ML18096B319
Person / Time
Site: Salem  PSEG icon.png
Issue date: 07/17/1992
From: Paluzzi V
Public Service Enterprise Group
To:
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ML18096B317 List:
References
PROC-920717, NUDOCS 9303090326
Download: ML18096B319 (95)
v • d • e


Text

.. *

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SALEM NUCLEAR GENERA'rING' *sT.A'.ttlION

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OFFSITE O()SE CALCULAT.!'ON M$'UAL ,  :

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Revision a July. *-1~9-~

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Approval S\)RC Chaimn:~~~*i Datil: 24u~i Mtq. #' *if,/,'-a.Al'

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9303090326 930226 PDR ADOCK 05000272 R PDR

Salem ODCM Rev~ 8 SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents Introduction . . . . -. - . . * . . . . . * *

  • _ * ........ 1 1.0 Liquid Effluents Lt Radiation Monitoring Instrumentation *and Controls . 2 1.2 Liquid Effluent Monitor Setpoint Determination 3 1.2.1 Liquid Effluent Monitors (Radwaste, steam Generator Blowdown and Service Water) * * . 4 1.2.2 Conservative Default Values * * * * * * * . s 1.3 L1quid Effluent Concentration Limits - 10 CFR 20 7 1.4 Liquid Effluent Dose Calculations - 10 CFR 50 . * . a 1.4.1 Member of the Public Dose - Liquid Effluents 8 1.4.2 Simplified Liquid Effluent Dose Calculation 10 1.5 Secondary Side Radioactive Liquid Effluents - Dose calculations During Primary to Secondary Leakage. . 11 1.6 Liquid Effluent Dose Projection * * * * * * . * . 13 I ,

2.0 Gaseous Effluents 2.1 Radiation Monitoring Instrumentation and Controls .

2.2 Gaseous Effluent Monitor Setpoint Determination . .

2.2.1 Containment and Plant Monitor 2.2.2 Conservative Default Values * * * * * . . .

2.3 Gaseous Effluent Instantaneous 15 17 17 19 Dose Rate Calculations - 10 CFR 20. * * * * * . 20 2.3.1 Site Boundary Dose Rate - Noble Gases. . 20

2. 3. 2 Site Boundary Dose Rate - .,

Radioiodine and Particulates * * * * * :i . . 21 2.4 Noble Gas Effluent Dose Calculations - 10 c::FRt,50 . . 24 2.4.1 UNRESTRICTED AREA Dose - Noble Gases * * . . 24 2.4.2 Simplified Dose Calculation for Noble Gases. 25 2.5 Radioiodine and Particulate Dose Calculations -

10 CFR 50. * * * * * * * * * * * * * * * * * . . . 26 2.5.1 UNRESTRICTED AREA Dose -

Radioiodine and Particulates * * * * * * . . 26 2.s.2 Simplified Dose Calculation for ~adioiodines and Particulates * * * * * * * * * * . . 27

-2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations * * * * * * * * * * * * . . . 28 2.7 Gaseous Effluent Dose Projection * * * * * * * . . 32 3.0 Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY. . 33 3.2 Doses to MEMBERS OF THE PUBLIC - 40 CFR 190 *

  • 34
3. 2 .1 Effluent Dose Calculations * * * * -* . . 35 3.2.2 Direct Exposure Determination * * * * .
  • 35 i
  • Table of Contents - continued 4.0 Salem ODCM Radiological Environmental Monitoring Program. .

4.1 Sampling Program . . * . * . . * * * * * * .

Rev. 8

  • .36
  • 36 4.2 Interlaboratory Comparison Program * * . * . * *
  • 37 Tables 1-1 Parameters for Liquid Alarm Setpoint Determination

- Unit 1 . . . . . . . . . . . . . . . . . . . . . 41 1-2 Parameters for Liquid Alarm Setpoint Determination

- Unit 2 . . . . . . . . . . . . . . . . . . . . . 42 1-3 Site Related Ingestion Dose Commitment Factors, Am * * * * * * * * * * * * * * * ~ ... 43-44 1-4 Bioaccumulation Factors (BFi) * * * * * * * . . . . 45 2-1 Dose Factors for Noble Gases * * * * * * * * . . . 48 2-2 Parameters for Gaseous Alarm Setpoint Determinations 2-3

- Unit 1 * * * * * * * ... * * * * * * * . * . . 49 Parameters for Gaseous Alarm Setpoint Determinations

- Unit 2 . . . . . . . . . . ._ . . . . .. . . . . . 50 2-4 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations * * * * * * . . . 51 2-5 Pathway Dose Parameters - Atmospheric Releases 52-63

  • A-1 A-2 B-1 B-2 Calculation of Effective MPC - Unit 1 * . . . . . A-5 Calculation of Effective MPC - Unit 2 * * * * *
  • A-6 Adult Dose Contributions Fish Pathways Unit 1 * * *
  • and Drinking water

. . .and Adult Dose Contributions Fish

. .Drinking Water B-5 Pathways Unit 2 *.* * * * * * * * .* * * * * :_'r..

  • B-5 C-5 Effective Dose Factors * * * * * * * * * * * ,, * . C-6 Appendices Appendix A - Evaluation of Conservative, .Default MPC Value for Liquid Effluents * * * * . . A-1 Appendix B - Technical Basis for Effective Dose Factors -

Liquid Radioactive Effluents * * * . . B-1 Appendix c - Technical Bases for Effective Dose Factors -

Gaseous Radioactive Effluents . . c-1

    • Appendix D - Radiological Environmental Monitoring Proqram - Sample Type, Location and Analysis * * * * * * * * * * * * * . D-1 ii

-. Salem ODCM SALEK NUCLEAR GENERATING STATION Rev. a OFFSITB DOSB CALCULATIOH MAHOAL Introduction The Salem Offsite Dose Calculation Manual (ODCM) describes the methodology and parameters used in: 1) the calculation of radioactive liquid and gaseous effluent monitoring instrumentation alarm/trip setpoints; and 2) the calculation of radioactive liquid and gaseous concentrations, dose rates and cumulative quarterly and yearly doses. The methodology stated in this manual is acceptable for use in demonstrating compliance with 10 CFR 20.106, 10 CFR so, Appendix I and 40 CFR 190.

  • More conservative calculation methods and/or conditions (e.g.,

location and/or exposure pathways) expected to yield higher computed doses than appropriate for the maximally exposed person may be assumed in the dose evaluations.

The ODCM will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations. Changes will be made to the ODCM calculation methodoloqies and parameters as is deemed necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR S0.36a and Appendix I for demonstrating radi~active effluents are ALARA.

NOTE.: As used throughout this document, excluding acronyms, words appearing all capitalized denote the application of definitions as used in the Salem Technical Specifications .

  • 1
  • Salem ODCM Rev. 8 1.0 Liquid Effluents 1.1 Radiation Monitoring Instrumentation and controls The liquid effluent monitoring instrumentation and .controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Salem Radiological Effluent Technical Specifications are summarized as follows: ,
1) Alarm Cand Automatic Termination> - l-Rl8 (Unit 1) and 2-Ris (Unit 2) provide the alarm and automatic termination of liquid radioactive material releases as requir~d by *

1-Rl9 A,B,C,and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R~9 A,B,C and D provide this function for Unit 2.

2) Alarm Conly> - The alarm functions for the Service ~ter system are provided by the radiation monitors on the*

Containment Fan Cooler discharges (1-R 13 A,B,C,D and E for Unit 1 and 2-R 13 A,B,and c for Unit 2).

Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored (with an alarm function) by R37 prior to release to the environment.

Liquid radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented as Figures 1-1 and 1-2 for Units l and 2, respectively .

  • 2

Salem ODCM Rev. 8 1.2 Liqui4 lffluent Monitor sotpoint Determination Per the requirements of Technical Specification 3.3.3.8, alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of Specification 3.11.1.1 are met (i.e., the concentration- of radioactive material released in liquid effluents to UNRESTRICTED AREAS.shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 2, for radionuclides and 2.oE-04* uCi/ml for dissolved or entrained noble gases). The f ollowinq equation* must be satisfied to meet the liquid effluent restrictions:

  • where:

cs C (F+f) f (1.1) c = the effluent concentration limit of ~chnical Specification (3.11.1.1) implementing the 10 CFR 20 MPC for the site, in uCi/ml .

c = the setpoint, in uCi/ml, of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release; the setpoint, represents a value which, if exceeded, would result in concentrations exceeding the limits of 10 CFR 20 in the UNRESTRICTED AREA f = the flow rate at the radiation monitor location, in volume per unit time, but in the same units as F, below F = the dilution water flow rate as measured prior to the release point, in volume per unit time

[Note that if no dilution is provided, cs c. Also, note that when (F) is large compared to (f), then (F + f) = F.]

Salem ODCM Rev. a Liquid Effluent Monitors CRadwaste, steam Generator Blowdown. Chemical Waste Basin and service water. The setpoints for the liquid effluent monitors at the Salem Nuclear Generating station are determined by the following equations:

MPCe

  • SEN
  • CW SP S, + bkq (1. 2)

RR with:

l: Ci MPCe = ------ (1. 3)

Ci

  • where:

SP = alarm setpoint correspondinq to the maximum allowable release rate (cpm)

MPCe = an effective MPC value for the mixture of *;:,

radionuclides in the effluent stream (uCi/mlL)

Ci = the concentration of radionuclide i in the undiluted liquid* effluents (uCi/ml)*

9NOTE: The concentration mix must include the most recent composite of alpha emitters, sr-89, sr-90, Fe-55, and H-3 per Technical Specification J.11.1.1.

MP Ci = the MPC value correspondinq to radionuclide i from 10 CFR 20, Appendix B, Table II, Column*2 (uCi/ml)

SEN = the sensitivity value to which the monitor is calibrated (cpm per uCi/ml)

  • cw = the circulating water flow rate (dilution water flow) at the time of release (qal/min)

RR = the liquid effluent release rate (qal/min) bkg = the backqround of the monitor (cpm)

  • 4
  • Salem ODCM Rev. a The radioactivity monitor setpoint equation (1.2) remains valid during outages when the circulating water dilution is potentially at its lowest value. Reduction of the waste stream flow (RR) may be necessary during these periods to meet the discharge criteria.

However, in order to maximize the available plant discharge dilution and thereby minimize the potential offsite doses, releases from either Unit-1 or Unit-2 may be routed to either the Unit-1 or Unit-2 Circulating Water System discharge. This routing is possible via interconnections between the Service Water Systems (see Figures 1 and 2). Procedural restrictions prevent simultaneous releases from either a single unit or both units into a single Circulating Water system

. discharge.

  • 1.2.2 conservative Default Values. Conserv~tive alarm setpoints may be determined through the use of default parameters. Tables 1-1 and 1-2 summarize all current default values in use for Salem Unit-1 and Unit-2, respectively. They are based upon th,~.:
  • -11 following:

a) substitution of the effective MPC value with a default value of 4.71E-06 uci/ml (Unit 1) and 3.38E-06 uci/ml (Unit 2). (refer to Appendix A for justification);

b) *or additional conservatism*, substit~tion of the I~131 MPC value of JE-07 uci/ml for the Ri9 steam Generator Slowdown Monitors, the R-37 Chemical waste Baain monitor and the R-13 Service Water Monitors.

  • Based upon the potential for I-131 to be present in the secondary and service water systems, the use of the default effective MPC value as derived in Appendix A may be non-conservative for the 1, 2 R-19 SGBD monitors, the R-37 Chemical Waste Basin Monitor and the R-13 Service Water
  • Monitors
  • s

-~

    • c) d)

Salem ODCM Rev. 8 substitutions of the operational circulating water flow with the lowest flow, in gal/min; and, substitutions of the effluent release rate with the highest allowed rate, in gal/min.

With pre-established alarm setpoints, it is possible to control the radwaste release rate (RR) to ensure the inequality of equation {1.2) is maintained under changing values for MPGe and for differing Circulating Water System dilutions

  • The Unit 2 Service Water system utilizes the Unit 1 Circulating Water-* systb for dilution prior to release to the river. It is possible to have the Unit 1 Circulating Water system. out of service when Unit 1 is in an outage. So, for conservatism no dilution is used for determining a 2R13 default alarm setpoint.

Because no dilution is considered and the 2R13 monitor sensitivity is high, the MPCe of 3.JSE-06 uCi/ml is used in calculating the alarm setpoint (otherwise using 3E-07 uCi/ml would result in an alarm setpoint of 1 cpm) *

  • 6
  • 1.3 Salem ODCM Liquit lffluent concentration Limits - 10 CPR 20 Rev. a Technical Specification 3.11.1.1 limits the concentration of radioactive material in liquid effluents (after dilution in the circulating Water System) to less than the concentrations as specified in 10 CFR 20, Appendix B, Table II, Column 2 for radionuclides other than noble gases. Noble gases are limited to a diluted concentration of 2.0E-04 uCi/ml. Release rates are controlled and radiation monitor alarm setpoints are established as addressed above to ensure that these concentration limits are not exceeded. However, in the event any liquid release results in an alarm setpoint being exceeded, an evaluation of compliance

Ci RR ~-

~

\~

E

  • -~ 1 (1.4)

~~ cw + RR where:

Ci = actual concentration of radionuclide i as measured in the undiluted liquid effluent (~Ci/ml)

~~ =- the MPC value corresponding to radionuclide i from 10 CFR 20, Appendix B, Table II, Column 2 (uCi/ml)

=- 2E-04 uCi/ml for dissolved or entrained noble gases RR = the actual liquid effluent release rate (gal/min) cw = the actual circulating water flow rate (dilution water flow) at the time of the release (gal/min)

  • 7

Salem ODCM Rev. a

~* 1.4 Liqui4 lffluent Dose Calculation - 10 Cl'R so 1.4.1 MEMBER OP THE PUBLIC Dose - Liquid Effluents.

Technical Specific~tion 3.11.1.2 limits the dose or dose commitment to MEMBERS OF THE PUBLIC from radioactive materials in liquid effluents from each unit of the Salem Nuclear Generating Station to:

- during any calendar quarter;

~ 105 mrem to total body per unit

~ 5.0 mrem to any organ per unit

- during any calendar year;

~ 3.0 mrem to total body per unit

~ 10.0 mrem to any organ per unit.

Per the surveillance requirements of Technical Specification 4.11.1.2, the following calculation methods shall be used for

  • determining the dose or dose commitment due to the liquid radioactive effluents from Salem.

l.67E-02

  • VOL

( 1. 9'l cw where:

dose or dose commitment to organ o (mrem). Total body dose can also be calculated using site- related total body dose commitment factor.

site-related ingestion dose commitment factor to the total body or any organ o for radionuclide i (mrem/hr per uCi/ml) -

  • average concentration of radionuclide i, in undiluted liquid effluent representative of the volume VOL *

(uCi/ml)

VOL= volume of liquid effluent released (gal) cw = average circulating water discharge rate during release period (gal/min)

1. 67E-02 = conversion factor (hr/min)
    • 8
  • Salem ODCM Rev. 8 The site-related ingestion dose/dose commitment factors (Aw) are presented in Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:

( 1. 6) where:

Aw = composite dose parameter for the total body or critical organ o of an adult for radionuclide i, for the fish and invertebrate ingestion pathways (mrem/hr per uci/ml)

UI = adult invertebrate consumption (5 kg/yr)

Bii = bioaccumulation factor for radionuclide*i in invertebrates from Table 1-4 (pCi/kg per pCi/l)

UF = adult fish consumption (21 kg/yr)

BFi = bioaccumulation factor for radionuclide i in fish from Table 1-4 (pCi/kg per pCi/l) dose conversion factor for nuclide i for adults in pre~

selected organ, o, from Table E-11 of Regulatory Guide 1.109 (mrem/pCi) -

1.14E+05= conversion factor (pCi/uCi

  • ml/kg per hr/yr)

The radionuclides included in the periodic dose assessmen~ per the requirements of Technical Specification 3/4.11.1.2 are.tho~e as identified by gamma spectral analysis of the liquid waste samples collected and analyzed per the requirements of Technical Specification 3/4.11.1.1, Table 4.11-1.

Radionuclid- requiring radiochemical analysis (e.g., Sr-89 and sr-

90) will be added to the dose analysts at a frequency consistent with the required minimum analysis frequency of Technical Specification Table 4.11-1 .
    • 9
  • Salem ODCM 1.4.2 Simplified Liquid Effluent Dose Calculation.

Rev. a In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following simplified dose calculation equation may be used for demonstrating compliance with the dose limits of Technical Specification 3.11.1.2. (Refer to Appendix B for the derivation and justification for this simplified method.)

Total Body 1.21E+03

  • VOL D111 = (1.7) cw

~ Maximum Organ 2.52E+04

  • VOL Dmu = (1. 8) cw where:

q = average concentration of radionuclide i, in undiluted liquid effluent representative of the volume VOL (uCi/ml)

VOL = volwaa of liquid effluent released (gal) cw -~

average circulating water dischar9e r~te during release period (gal/min)

  • Dlb o_ ----- conservatively evaluated total body dose (mrem) conservatively evaluated maximum organ dose (mrem)
1. 21E+03 = conversion factor (hr/min) and the conservative total body dose conversion factor (Fe-59, total body -- 7.27E+04 mrem/hr per uCi/ml) 2.52E+04 = conversion factor (hr/min) and the conservative maximum organ dose conversion factor (Nb-95, GI-LLI

-- 1.51E+06 mrem/hr per uCi/ml)

  • 10
  • Salem ODCM 1.s secondary Sid* Radioactive Liquid Effluents and Dose Calculat~ODI Qurinq Primary to Secondary Leakage Rev. 8 During periods of primary to secondary leakage (i.e., steam generator tube leaks), radioactive material will be transmitted from the primary system to the secondary system. The potential exists for the release of radioactive material to the off-site environment (Delaware' River) via secondary system dischargeso Potentially significant radioactive material levels and potential releases are controlled/monitored by the Stea~ Generator blowdown monitors (Rl9) and the Chemical Waste Basin monitor (R37). However to ensure compliance with the regulatory limits on radioactive material releases, it may be desirable to account for potential releases from the secondary system during periods of primary to secondary leakage. Any potentially significant releases will be via the Chemical Waste Basin with the major source of*~activity

. '((.

being the Steam Generator blowdown.

With identified radioactive material levels in the secondary system, appropriate samples should be collected and analyzed for the principal gamma emitting radionuclides~ Based on the identified radioactive material levels and the volume of water discharged, the resulting environmental doses may be calculated based on equation (1.5) .

    • 11

Salem ODCM Rev. a Because the release rate from the secondary system is indirect (e.g. , SG blowdown is normally routed to condenser where the condensate clean-up system will remove much of the radioactive material), samples should be collected from the final release point (i.e., Chemical Waste Basin) for quantifying the radioactive material releases. However, for conservatism and ease of controlling and quantifying all potential release paths, it is prudent to sample the SG blowdown and to assume all radioactive material is released directly to the environment via the Chemical waste Basin. This approach while not exact, is conservative and ensures timely analysis for regulatory compliance. Accounting for radioactive material retention of the condensate clean-up system

  • ion exchange resins may be needed to more accurately account for actual releases *
  • 12

Salem ODCM Rev. a 1.6 Liqui4 lffluent Dose Pro1ections Technical Specification 3.11.1.3 requires that the liquid radioactive waste processing system be used to reduce the radioactive material levels in the liquid waste prior to release when the quarterly projected doses exceed:

0.375 mrem to the total body, or 1.25 mrem to any organ.

The* applicable liquid waste processing system for maintaining radioactive material releases ALARA is the ion exchange system as delineated in Figure 1-3. Alternately, the waste evaporator as presented in the Salem FSAR.has processing capabilities meeting the

  • NRC ALARA design requirements and may be used in conjunction or in lieu of the ion exchange system for waste processing requirements in accordance with Technical Specification 3.11.1.3. These processing requirements are applicable to each unit indiV'tidually.
  • i.

Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing requirements for the other unit *

    • 13

-* Salem ODCM Dose projections are made at least once per 31 days following equations:

Rev. 8 by the Dtbp = Dlb (91 I d) (1. 9)

Dmaxp = Dmu (91 I d) ( 1. 10) where:

Dtbp = the total body dose projection for current calendar quarter (mrem) *

  • Dtb = the total body dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) om.up = the maximum organ dose projection for current calendar quarter (mrem)

Dmu = the maximum organ dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) d = the number of days to date for current calendar quarter 91 = the number of days in a calendar quarter

  • 14
  • 2.0 2.1 Gas90u1 Bffluents Salem ODCM Radiation Monitoring Instrumentation and Controls Rev. a The gaseous effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Radiological Effluent Technical Specifications are summarized as follows:
1) Waste Gas Holdup System - The vent header gases are collected by the waste gas holdup system. Gases may be recycled to provide cover gas for the eves hold-up tank or held in the waste gas tanks for decay prior to release. waste gas decay tanks are batch *released after sampling and analysis. The tanks are discharged via the Plant Vent.

l-R41C provides noble gas monitoring and automatic isolation of waste gas decay tank releases for Unit-1. This function is provided by 2-R41C for Unit-2.

2) Containment Purge and Pressure/Vacuum Relief - containment purges and pressure/vacuum reliefs are released to the atmosphere via the respective unit Plant Vent. Noble gas monitoring and auto isolation function are provided by l-R41C for Unit-1 and 2-R41C for Unit-2. Additionally, in accordance with Technical Specification 3. 3. 3. 9, Table 3. 3-13, 1-Rl2A and 2-Rl2A may be used to provide the containment monitoring and automatic isolation function during purge. and pressu~e/vacuum reliefs.* *  :\:

?1

3) Plant Vent - The Plant Vent for each respective unit receives discharges from the waste gas hold-up system, condenser evacuation system, containment purge and pressure/vacuum reliefs, and the Auxiliary Building ventilation. Effluents are monitored by R41C, a flow through gross activity monitor (for noble gas monitoring).

Additionally, in-line gross activity monitors, (1-R16 and

  • The R12A monitors also provide the safety function of containment isolation in the event of a fuel handling accident during refueling. During MODE 6 in accordance with Technical Specification 3/4.3.3, Table 3.3-6, the R12A alarm/trip setpoint shall be established at twice background, providing early indication and containment isolation accompanying unexpected increases in containment airborne radioactive material levels indicative of a fuel degradation. The R41C monitor may also provide
    • this function if the R12A monitor is inoperable during MODE 6 .

15

  • 3)

Salem ODCM Rev. 8 Plant Vent Ccont'dl Rl6) provide redundant back-up monitoring capabilities to the R41C monitors. Radioiodine and particulate sampling capabilities are provided by charcoal cartridge and filter medium samplers with redundant back-up sampling capabilities provided by R41B and R41A, respectively. Plant Vent flow rate is measured and as a back-up may be determined empirically as a function of fan operation (fan curves). Sampler flow rates are determined by flow rate instrumentation (e.g., venturi rotameter) .

Gaseous radioactive waste flow diagrams with the applicable, associated radiation monitoring instrumentation and controls are presented as Figures 2-1 and 2-2 for Units 1 and 2, respectively .

  • 16
  • 2.2
2. 2. 1 Gaseoq1 Effluent Monitor setpoint Determination containment and Plant vent Monitor.

Salem ODCM Rev.

Per the requirements of a

Technical Specification 3.3.3.9, alarm setpoints shall be established for the gaseous effluent monitoring instrumentation to ensure that the release rate of noble gases does not exceed the limits of Specification 3.11.2.1, which corresponds to a dose rate at the SITE BOUNDARY of 500 mrem/year to the total body or 3000 mrem/year to the skin. Based on a grab sample analysis of the applicable release (i.e., grab sample of the Containment atmosphere, waste gas decay tank, or Plant Vent), the radiation monitoring alarm setpoints may be established by the following calculation method. The measured

. radionuclide concentrations and

  • release rate are used to calculate the frac~ion of the allowable release rate, as limited by Specification J.11.2.1, by the equation:

FRAC = [ 4. 72E+02

  • X/Q
  • l: (Ci * ~) ] / 500 ( 2. 1)

FRAC = [4.72E+02

  • X/Q *VF* I: {Ci*{~+ 1.1 ~))] / 3000 (2.2) where FRAC = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate .

X/Q = annual average meteorological dispersion 3to the controlling site boundary location {sec/m )

VF = ventilation system flow rate for the applicable release point and monitor (ft3 /min)

= concentration of noble gas radionuclide i as determine radioanalysis of grab sample {uCi/cm3)

= total body dose conversion factor 3 for noble gas radionuclide i (mrem/yr per uci/m from Table 2-1)

= beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m3 from Table 2-1)

  • 17
  • =

Salem ODCM gamma air dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m3 from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad)

Rev. a 500 = total body dose rate limit (mrem/yr) 3000 = skin dose rate limit (mrem/yr) 4.72 E+02 = conversion factor (cm /ft3

  • min/sec)

Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoints for the applicable monitors (R16, R41C, and/or R12A) may be calculated by the equation:

SP = [AF

  • I: Ci
  • SEN / FRAC] + bkg ( 2. 3) where:

SP = alarm setpoint corresponding to the maximum allowable release rate (cpm)

SEN = monitor sensitivity (cpm per uCi/cm3 )

bkg = background of the monitor (cpm)

AF = administrative allocation factor for the specific monitor and type release, which corresponds to the fraction of the total allowable release rate that is administratively allocated to the release.

The allocation factor (AF) is an administrative control i~posed to

,...\;

ensure that combined releases from Salem Uni ts 1 and 2 and Hope Creek will not exceed the regulatory limits on release rate from the site {i.e., the release rate limits of Technical Specification 3.11.2.1). Normally, the combined AF value for Salem Units 1 and 2 is equal to o

  • 5 ( o. 25 per unit) , with the
  • remainder o. 5 allocated to Hope Creek. Any increase in AF above 0.5 for the Salem Nuclear Generating station will be coordinated with the Hope Creek Generating Station to ensure that the combined allocation factors for all units do not exceed 1.0 *
    • 18
  • 2.2.2 can be Cog1eryative Default Values.

established, in lieu of the Salem ODCM individual Rev. a A conservative alarm setpoint radionuclide evaluation based on the grab sample analysis, to eliminate the potential of periodically having to adjust the setpoint to reflect minor changes in radionuclide distribution and variations in release flow rate. The alarm setpoint may be conservatively determined by the default values presented in Table 2-2 and 2-3 for Units 1 and 2, respectively. These values are based upon:

the maximum ventilation (or purge) flow rate; a radionuclide distribution* comprised of 95% Xe-133, 2%

Xe-135, 1% Xe-133m, 1% Kr-88 and 1% Kr-85; and

- an administrative allocation factor of 0.25 to.conservatively ensure that any simultaneous releases from Salem Units 1 and 2 do not exceed the maximum allowable release rate.

  • For this radionuclide distribution, the alarm setpoint based oh the total body dose rate is more restrictive than the corresponding setpoint based on the skin dose rate. The resulting conservative,

..:~

.'{\-,:

default setpoints are presented in Tables 2-2 and 2-3.  :.;.

Adopted from ANSI N237-1976/ANS-18.1, Source Term Specif_ications, Table 6

  • 19
  • 2.3 2.3.1 Boundary Dose Rate Salem ODCM GastOg* lffluent Instantaneous Dose Rate Calculations -

10 CD 20 sit* Noble Gases.

Rev. 8 Technical Specification 3.11.2.la limits the dose rate at the SITE BOUNDARY due to noble gas releases to ~500 mrem/yr, total body and ~3000 mrem/yr, skin. Radiation monitor alarm setpoints are established to ensure that these release limits are not exceeded. In the event any gaseous releases from the station results in an alarm setpoint being exceeded, an evaluation of the SITE BOUNDARY dose rate resulting from the release shall be performed using the following equations:

( 2. 4) and D1 = X/Q

  • l: ( (~ + 1.1~)
  • Qi) (2. 5) where:

= total body dose rate (mrem/yr) .,

= skin dose rate (mrem/yr) _}

= atmospheric dispersion to3 the controlling SITE BOUNDARY location (sec/m )

= average release rate of radionuclide i over the release_ period under evaluation (uCi/sec)

= total body dose conversion factor for noble gas radionuclide i (mrem/yr per uci/m3 , from Table 2-1)

= beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m3 , :fro~ Table 2-1)

- q&J11Da air dose conversion factor for noble gas radionuclide i (mrad/yr per uci/m3 , from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad)

As appropriate, simultaneous releases from sa_lem Units 1 and 2 and Hope creek will be considered in evaluating compliance with the releas*e rate limits of Specification 3 .11. 2. la, following any

  • 20
  • Salem ODCM release exceeding the above prescribed alarm setpoints.

Rev. a Monitor indications {readings) may be averaged over a time period not to exceed 15 minutes when determining noble gas release rate based on correlation of the monitor reading and monitor sensitivity. The 15 minute averaging is needed to allow for reasonable monitor response to potentially changing radioactive material concentrations and to exclude potential electronic spikes in monitor readings that may be unrelated to radioactive material releases. As identified, any electronic spiking monitor responses may be excluded from the analysis.

NOTE: For administrative purposes, more conservative alarm setpoints than those as prescribed above may be imposed.

However, conditions exceeding these more limiting alarm setpoints do not necessarily indicate radioactive material release rates exceeding the limits of Technical Specification 3.11.2.la. Provided actual releases do not result in radiation monitor indications exceeding alarm setpoint values based on the above criteria, no further analyses are required for demonstrating compliance with the limits of Specification 3.11.2.la.

.l:

Actual meteorological conditions concurrent with the relea~e period or the default, annual average dispersion parameters as presented in Table 2-4 may be used for evaluating the gaseous effluent dose rate.

Sit* Boundary Dose Rate - Radioiodine and Particulates.

Technical Specification 3.11.2.1.b limits the dose rate to ~1500 lives greater than 8 days. To demonstrate compliance with this

  • 21
  • Salem ODCM Rev. 8 limit, an evaluation is performed at a frequency no greater than that corresponding to the sampling and analysis time period (e.g.,

nominally once per 7 days). The following equation shall be used for the dose rate evaluation:

(2. 6) where:

average organ dose rate over the sampling time period (mrem/yr)

X/Q = atmospheric dispersion to the controlling* SITE BOUNDARY location for the inhalation pathway (sec/m3 )

~o = dose parameter for radionuclide i (mrem/yr per uCi/m3 )

and organ o for the child inhalation pathway from Table 2-5 average release rate over the appropriate sampling period and analysis frequency for radionuclide i --

I-131, I-133, ttitium or other radionuclide in particulate form with half-life greater than 8 days (uCi/sec)

By substituting 1500 mrem/yr for D0 and solving for Qu an allowable release rate for I-131 can be determined. Based on th*. annual average meteorological dispersion (see Table 2-4) and the most limiting potential pathway, age group and organ (inhalation, child, thyroid -- Rm= l.62E+07.mrem/yr per uCi/m3 ), the allowable release

  • rate for I-131 is 42 uCi/sec. Reducing this:release rate by a factor of 4 to account for potential dose contributions from other radioactive particulate material and other release points (e.g.,

Hope Creek), the corresponding release rate allocated to each of

  • 22
  • the Sal.em units is 10.5 uCi/sec.

Salem ODCM Rev. s For a 7 day period, which is the nominal sampling and analysis frequency for I-131, the cumulative release is 6.3 Ci. Therefore, as long as the I-131 releases in any 7 day period do not exceed 6. 3 Ci, no additional analyses are needed for verifying compliance with the Technical Specification 3.11.2.1.b limits on allowable release rate *

  • 23
  • 2.4 trlfBISTRICTBD AREA Dose Salem ODCM Noble Gaa Effluent Dose Calculations - 10 CFR so Noble Gases.

Rev. 8 Technical Specification 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate compliance with the quarterly* dose limits of ss mrad, gamma-air and SlO mrad, beta-air and the calendar year limits SlO mrad, gamma-air and s20 mrad, beta-air.

The limits are applicable separately to each unit and are not combined site limits. The following equations shall be used to calculate the gamma-air and beta-air doses:

= 3.17E-08

  • X/Q
  • E {~ * ~) (2.7) and
  • where:

=

= 3.17E-08

  • X/Q
  • E {~ * ~)

air dose due to gamma emissions for noble g*(\ls

( 2_. 8) radionuclides (mrad) ,,\:

= air dose due to beta emissions for noble gas radionuclides (mrad)

X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location {sec/m3) cumulative release of noble gas radionuclide i over the period of interest {uCi) where uci = {uci/cc} (cc released) or (uCi/sec) (sec released) air dose factor due to gamma emis~ions. from noble gas radionuclide i (mrad/yr per uCi/m3 , from Table 2-1) air dose factor due to beta emissions from noble gas radionuclide i (mrad/yr per uCi/m3 , Table 2-1) 3 .17E-08 = conversion factor (yr/sec)

  • 24
  • 2.4.2 Sipplifie4 Dose Calculation for Noble Gases.

Salem ODCM individual noble gas radionuclide dose assessment as presented Rev. 8 In lieu of the above, the following simplified dose calculation equations shall be used for verifying compliance with the dose limits of Technical Specification J.11.2.2. (Refer to Appendix c for the derivation and justification for this simplified method.)

J.17E-08 o, = (2. 9) 0.50 and J.17E-08

=

  • X/Q
  • Neff
  • l: Qi (2 .10)
o. 50 v

where:

MetT = 5.JE+02, effective gamma-air dose factor (mrad/yr per uCi/m3 )

1. lE+OJ, effective beta-air dose factor ..(mrad/yr per uCi/m3 )  : 'l cumulative release for all noble gas radibnuclides (uCi) where uCi = * (uCi/cc) (cc released) or (uCi/sec) (sec released) 0.50 = conservatism factor to account for potential variability in the radionuclide distribution Actµal meteorological conditions concurrent witQ the, release period or the default, annual average dispersion parameters as presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doses *
  • 25

Salem ODCM Rev. a

  • .2.5 Radioiodin* and Particulate Dose Calculations - 10 CFR so 2.s.1 tJNBBSTBICTED AREA Dose - Radioiodine and Particulates.

In accordance with reqllirements of Technical Specification 3 .11. 2. 3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit of ~7.5 mrem and calendar year limit ~15 mrem to any organ. The following equation shall be used to evaluate the maximum organ dose due to releases of I-131, tritium and particulates with half-lives greater than 8 days:

Daop = 3. l 7E-08

  • W
  • SFp * :t (Rq,
  • Qi) (2.11) where:

Daop = dose or dose commitment via all pathways p and controlling age group a {as identified in Table 2-4) to organ *O, including the total body {mrem) w = atmospheric dispersion parameter to the controlling location{s) as ide~tified in Table 2-4

  • X/Q D/Q

=

=

atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways {sec/m3)

  • atmospheric deposition for vegetation, milk and ground plane exposure pathways (m-2 )

Ru,, = dose factor for radionuclide i {mrem/yr per ~i/m3 ) or

{m2 - mrem/yr per uCi/sec) and organ o from Table 2-s for each age group and the applicable pathway p as identified in Table 2-4. Values for Rm were derived in accordance with the methods described in NUREG-0133

= cumulative release over the period of interest for radionuclide i -- I-131 or radioactive material in particulate form with half-life g~eater than 8 days (uCi). - .

= annual seasonal correction factor to account for the fraction of the year that the applicable exposure pathway does not exist.

1) For milk and vegetation exposure pathways:

= A six month fresh vegetation and grazing season. (May through October)

= 0.5

2) For inhalation and ground plane exposure pathways:

= 1.0 26

  • Salem ODCM For evaluating the maximum exposed individual, the infant age group is controlling for the milk pathway.

Rev. 8 Only the controlling age group as identified in Table 2-4 need be evaluated for compliance with Technical Specification 3.11.2.3.

2.s.2 simplified Dose Calculation for Radioiodines and Particulates. In lieu of the individual radionuclide (I-131 and particulates) dose assessment as presented above, the following simplified dose calculation equation may be used for verifying compliance with the dose limits of Technical Specification 3. 11. 2. 3 (refer to Appendix D for the derivation and justification of this simplified method).

Dmu = 3 . 17E-08

  • W
  • SFp
  • Rr.131
  • I: Qi ( 2. 12) where:

maximum organ dose (mrem)

I-131 dose parameter for the thyroid for the identified controlling pathway

= 1.05E+12, infant thyroid dose parameter with the cow-milk pathway controlling (m2 - mrem/y~ per uci/sec) ,Y D/Q for radioiodine, 2.lE-10 1/m2

  • cumulative release over the period of interest for radionuclide I -- I-131 or radioactive material in particulate from with half life greater than a days (uCi)

The location of exposure pathways and the maximum organ dose

. :;i calculation may be based on the available pathways in the surrounding environment of Salem as identified by the annual land-use census (Technical Specification 3.12.2). Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4 .

  • 27
  • 2.6 secondary Calculations Side Radioactive Gaseous Salem ODCM Effluents Rev. 8 and Dose During periods of primary.to secondary leakage, minor levels of radioactive material may be released via the secondary system to the atmosphere. Non-condensables (e.g., noble gases) will be predominately released via the condenser evacuation system and will*

be monitored and quantified by the routine plant vent monitoring and sampling system and procedures (e.g., RlS on condenser evacuation, R41C on plant vent, and the plant vent particulate and charcoal samplers).

However, if the Steam Generator blowdown is routed directly to the

  • Chemical Waste Basin (via the SG blowdown flash tank) instead of being recycled through the condenser, it may be desirable to account for the potential atmospheric releases of radioiodines and particulates from the flash tank vent (i.e., release~; due to t:

moisture carry over)

  • Since this pathway is not sampled or monitored, it is necessary to calculate potential releases.

Based on the guidance in NRC NUREG-0133, tl}e releases of the radioiodinas and particulates shall be calculated by the equation:

(2.13) where:

Qi = the release rate of radionuclide, i, from the steam generator flash tank vent (uCi/sec)

    • 28
  • Ci

~b =

=

Salem ODCM Rev. 8 the concentration of radionuclide, i, in the secondary coolant water averaged over not more than one week (uCi/ml) the steam generator blowdown rate to the flash tank (ml/sec)

Fft = the fraction of blowdown flashed in the tank determined from a heat balance taken around the flash tank at the applicable reactor power level

  • SQ!tv = the measured steam quality in the flash tank vent; or an assumed value of 0.85, based on NUREG-0017.

Tritium releases via the steam flashing may also be quantified using the above equation with the assumption of a steam quality (SQ1tv) equal to o. Since the H-3 will be associated with the water molecules, it is not necessary to account for the moisture carry-over which is the transport media for the radioiodines and particulates.

  • Based on the design and operating conditions at Salem, the fraction of blowdown converted to steam (Fft) is approximately o. 48. The equa~ion simplifies to the following:

(2 .14)

For H-3, the simplified equation is:

Also durinq reactor shutdown operations with a radioactively contaminated secondary system, radioactive material may be released to the atmosphere via the atmospheric reliefs (PORV) and the safety

  • 29
  • Salem ODCM Rev. a reliefs on the main steam lines and via the steam driven auxiliary feed pump exhaust. The evaluation of the radioactive material concentration in the steam relative to that in the steam generator water is based on the guidance of NUREG-0017, Revision 1. The partitioning factors for the radioiodines is 0.01 and is 0.001 for all other particulate radioactive material. The resulting equation for quantifying releases via the atmospheric steam releases is:

(2 .16) where:

Qij = release rate of radionuclide i via pathway j {uCi/sec) cij = concentration of radionuclide i, in pathway j, (uCi/sec)

SFj = steam flow for.release pathway j

= 450,000 lb/hr per PORV .

= 800,000 lb/hr per safety relief valve

= 50,000 lb/hr for auxiliary feed pump exhaust PFi = partitioning factor, ratio of concentration in steam to that in the water in the steam generator

= 0.01 for radioiodines

= 0.005 for all other particulates

= 1. o for H-3 0.13= conversion factor - [(hr*ml) / (sec*lb)]

Any significant releases of noble gases via the atmospheric steam releases can be quantified in accordance with the calculation methods o~ the Salem Emergency Plan Implementation* Procedure .

  • 30

-* Salem ODCM Rev. 8 Alternately, the quantification of the release rate and cumulative releases may be based on actual samples of main steam collected at the R4 6 sample locations. The measured radionuclide concentration in the steam may be used for quantifyinq the noble gases, radioiodine and particulate releases.

Note: The expected mode of operation would be to isolate the effected steam generator, thereby reducinq the potential releases durinq the shutdown/cooldown process. Use of the above calculation methods should consider actual operating conditions and release mechanisms.

The calculated quantities of radioactive materials may be used as inputs to the equation (2.11) or (2.12) to calculate offsite doses for demonstratinq compliance with the Radioloqical Effluent Technical Specifications *

  • 31
  • 2.7 Gaseou* Bffluent Dose Proiection Salem ODCM Rev. 8 Technical Specification 3 .11. 2. 4 requires that the GASEOUS RADWASTE TREATMENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when projected doses exceed one-half the annual design objective rate in any calendar quarter, i.e., exceeding:

0.625 mrad/quarter, gamma air; 1.25 mrad/quarter, beta air; or 1.875 mrem/quarter, maximum organ.

The applicable gaseous processing systems for maintaining radioactive material releases ALARA are the Auxiliary Building normal ventilation system (filtration systems # 1,2 and 3) and the Waste Gas Decay Tanks as delineated in Figures 2-3 and 2-4 *

  • Dose projections are performed at least once per 31 days by. the following equations:

DIP = o, * (91 I d) (2.17)

Dtip = Db * (91 I d) (2 *:Ja>

Dmap = Dmu * (91 I d) (2 .19) where:

DIP = gamma air dose projection for current calendar quarter (mrad)

D, = gamma. air dose to date for current calendar quarter as determined by Equation 2.7 or ~.9 (mrem)

Dtip -- beta air dose projection for current calendar quarter

~ -

Dmup =

(mrad) beta air dose to date for current calendar quarter as determined by Equation 2.8 or 2.10 (mrem) maximum organ dose projection for current calendar quarter (mrem)

Dmu = maximum organ dose to date for current calendar quarter as determined by Equation 2.11 or 2.12 (mrem) d = number of days to date in current calendar quarter

    • 91 = number of days in a calendar quarter 32
  • 3.0 3.1 Special Dose Analyses Salem ODCM Doses Due To Activities Inside the SITE BOUNDARY Rev. a In accordance with Technical Specification 6. 9 .1.11, the Radioactive Effluent Release Report (RERR) submitted within 60 days after January 1 of each year shall include an assessm,ent of radiation doses from radioactive liquid and qaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.

The calculation methods as presented in Sections 2.4 and 2.5 may be used for determining the maximum potential dose to a MEMBER OF THE PUBLIC based on the parameters from Table 2-4

  • The default value for the meteorological dispersion data as presented in Table 2-3 may be used if current year meteorology is unavailable at .;the time

. l-of NRC reporting. However, a follow-up evaluation 'shall be performed when the data becomes available *

    • 33
  • 3.2 Salem ODCM Total dose to MEMBERS OP THE PUBLIC - 40 CJ'R 190 Rev. a The Radioactive Effluent Release Report (RERR) submitted within 60 days after January 1 of each year shall also include an assessment of the radiation dose to the likely most exposed MEMBER OF THE PUBLIC for reactor releases and other nearby uranium fuel cycle sources (including dose contributions from. effluents and direct radiation from on-site sources). For the likely most exposed MEMBER OF THE PUBLIC in the vicinity of Artificial Island, the sources of exposure need only consider the Salem Nuclear Generating Station and the Hope creek Nuclear Generating Station: No other
  • fuel cycle facilities contribute to the MEMBER OF THE PUBLIC dose for the Artificial Island vicinity.

The dose contribution from the operation of Hope Creek,}Nuclear Generating Station will be estimated based on the methods as presented in the Hope Creek Offsite Dose Calculation Manual (HCGS ODCM).

As appropriate for demonstrating/evaluating compliance with the limits of Technical Specification 3.11.4 (40 CFR 190), the results of the environmental monitoring program may be used for providing data on actual measured levels of radioactive material in the actual pathways of exposure .

  • 34
  • 3. 2 .1 Bfflutnt Dost Calculations.

Salem ODCM Rev. a For purposes of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.11 (RERR), dose calculations for the Salem Nuclear Generating Station may be performed using the calculation methods contained within this .ODCM; the conservative controlling pathways and locations of Table 2-4 or the actual pathways and locations as identified by the land use census (Technical Specification 3/4.12.2) may be used. Average annual meteorological dispersion parameters or meteorological conditions concurrent with the release period under evaluation may be used.

l.2.2 Direct Exposure Dose Determination. Any potentially

  • significant direct exposure contribution to. off-site individual doses may be evaluated based on the results of the environmental measurements (e.g., TLD, ion chamber measurements) and/or by the use of a radiation transport and shielding calculation} method.

Only during atypical conditions will there exist any potential for significant on-site sources at Salem that wouid yield potentially significant off-site doses (i.e., in excess of 1 mrem per year to a MEMBER OP THE PUBLIC) , that would require detailed evaluation for demon*stratinC) compliance with 40 CFR 190. However, should a situation exist whereby the direct exposure contribution is potentially significant, on-site measurements, off-site measurements and/or calculation techniques will be used for determination of dose for assessing 40 CFR 190 compliance .

  • 35
  • 4.0 Salem ODCM Radiological Environmental Monitoring Program Rev. 8 4.1 Sampling Program The operational phase of the Radiological Environmental Monitoring Program (REMP) is conducted in accordance with the requirements of Appendix A Technical Specification 3.12. The objectives of the program are:

- To determine whether any significant increases occur in the concentration of radionuclides in the critical pathways of exposure in the vicinity of Artificial Island; To determine if the operation of the Salem Nuclear Generating Stations has resulted in any increase in the inventory of long lived radionuclides in the environment;

  • To detect any changes in the ambient gamma radiation levels; and To verify that SNGS operations have no detrimental effects on the health and safety of the public or on the environmen~.

The sampling requirements (type of samples*, collection frequency and analysis) and sample locations are presented in Appendix E.

  • uoTB: No public drinking water samples or irrigation water samples are taken as these pathways are not directly effected by liquid effluents discharged from Salem Generating Station *
  • 36
  • 4.2 InterlaJ2oratory comparison Program Salem ODCM Rev.

Technical Specification 3.12.3 requires analyses be performed on a

radioactive material supplied as part of an Interlaboratory Comparison. Participation in an approved Interlaboratory Comparison Program provides a check on the preciseness of measurements of radioactive materials in environmental samples. A summary of the Interlaboratory Comparison Proqram results.will be provided in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.10 *

  • 37

RAOIATION MONITORING LIQUID RELEASES UNIT l FIGURE 1-1 u*-* c&£L

.I *~ *1 i::,

J. .. ..... Lr ..,.,.

  • I
~ i * '*; ...

..~

~- ........ .-".:Jlfl& ,

dBftfu

-*----- '1 lo--

"IL"ll-w

-=wt '-- 111(111' J"" ..... r . . . . r &.1*r r-- ...... Ilk I* '* I*

. 1.,...

-m m-IDI

..." .. CllllUIMlll 111P1 1111111

_J *****---..

auau:aa

. ~,-:._...;

. *.~ *.

  • RADIATION ~ONITORING *** LIQUID RELEASES UNIT 2 FIGURE 1-2 u,_ .. ~&.

I dl&Ma

  • r..1, 1 ~
  • .IL.'WI. ~

....* :JI._.. ...... J' Ha m -~ er .....

  • r mlll'\.. .... -u&a'Tll*

w

.-m,_,..

~ .._

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  • tm

.... . r . . . . r ...., r-- -&..-

I* I* ** *I

  • a I

.... -----

  • P*

m-..

!Ml' lf**-**11

_J '

.. Cl&llJlllllll .... llllP

  • llOTE: Evaporator p.cltage and/or 1'.aMste delltneraltierr s1ste11 1'0 *****

AC&:l..IULAl'Ull U11&1*a

    • r.:a.. U*IMl*M (lllC) 0 I

I I

I I --*... -

    • ac.* COO&.* ...

--- -*-- :I --

.... c ...................

ftg 1-l S.I*

a.....u.a """"'-*** W.ar i;,.,....,

IU FOi llFCNllATIOI aw a"'"" rt *11,

  • Table 1-1 Salem ODCM Rev. 8 Parameters for Liquid Alarm Setpoint Determinations Unit 1 Parameter Actwl Defailt ~it* ec-nta Value Value MPC 0 calculated 4.71E-06
  • uci/ml calculated for each batch to be released.

MPC I-131 3.0E-07 N/A uci/11l I-131 MPC conservatively used for SG blowdown and Service Water monitor set....,ints.

c, measured N/A uci/ml taken froia g1111111a spectral analysis of l fm.iid effluent.

14P(1 as det emi necl N/A uci/11l taken froia 10 CFR 20, Appendix B, Table II. Col 2.

Sen 1-R18 as determined 2.9E+07 Cplll per Radwaate Effluent (Cs-137) uci/*l 1-R19 2.9E+07 Ste.. Generator &lowdown ccs-137)

(A,B,C,D) 1-R13 (A) 1-R13 CB) 1-R13 (C,D,E) 5.62E+07 5.98E+07 1.01E+08 Service Water - Contai1111ent Fan Cooling ccs-137)

I cw BS deten1inecl 1.00E+OS 9PI circulating ...ter *sys.tern - single cw m..a .,

RR 1-R18 as detenained 120 aim detenained prior to n;l~ase:

release rate can be ad-justed for 1-R19 Technical Specification c~liance 1-R13 120 Ste.. Generator blowdown rate per Generator Service Water flow rate for 2500 Contai1111ent fan coolers Setpoint 1-R11 calculated 1.13E+05(+bkg) Cplll Default alana setpoints: more conser:vative*velues may be used as 1-R19 - 7.25E+03C+blcg) deenied appropriate and desiralbe for ensuring regulatory c~liance 1-R13 CA) .. 6.70E+02(+bkg) and for .. intaining releases ALARA.

1-R13 (B) ..., 7.10E+02(+blcg) 1-R13 (C,D,E)- 1.09E+03(+bkg)

  • Refer to Appendix .A for derivation
    • The MPC*value of I-131 C3E-07 uci/ml) has been used for derivation of R19 Ste.. Generator blowdown and R13 Service Water monitor setpoints as discussed in Section 1.Z.Z 41
  • Table 1-2 Parameters for Liquid Alarm setpoint Determinations Unit 2 Salem ODCM Rev. 8 Parameter Actual Default *units Conmen ts Value Value MPC0 calculated 3.38E-06
  • uci/111l calculated for each batch to be released.

MPC 1*131 3.0E-07 N/A uci/ml 1*131 MPC conservatively used for SG blowdown, Service Water and Chemical Waste Basin monitor setnoints.

c. measured N/A uci/11l tak1n frCllll 98111111! spectral analysis of ll,.*id effluent.

MPC. as determined N/A uci /Ill taken frOll 10 CFR 20, Appendix B, Table II - Col. 2 Sen 2*R18 as determined 1.14E+08 cpnt/uc i /111l Redweate Effluent ccs-137) 2*R19CA,C) 1.26E+08 Ste.. Generator Blowdown (Cs-137) 2*R19(B) 1. 14E+08 2*R19(D) 1. 13E+08 2*R13 9.05E+07 Service Water

  • Containnent Fan Cooling (Ca-137>

.,_ R37 1.24E+08 C!'e11ical Waste Basin discharge cw as determined 1.0E+05 gpm Circulating Water System, single CW

~ (Note: no CW ~ in service for 2R13 monitor* see section 1.2.2)

RR 2*R18 aa deten1i necl 120 9fm deten1inecl prior to r~ease: release rate can be adjusted.:lr Technical Specification Ccq:il i *: e

  • 2*R19 120 Ste.. Generator Slowdown rate per Generator Service Weter flow rate for 2*R13 2500 Contairaent fan coolers R37 1200 Chemical Waste Baain discharge Setpoint 2-R1a calculated 3.20E5 C+bkg) Cpnl Default alarm setpoints: more conservative values may be used as 2-R19 - 2. 10E4 deellld appropriate and desirable for
  • CA,B,C,aJ ensuring regulatory ccq:iliance and for

..intaining releases ALARA.

2*R13 " 3.05E2 (+bkg)

R37" 3. 10E3 (+bkg)

  • Refer to Appendix A for derivation
    • The MPC value of 1*131 (3.0E-7 ucilml) has been used for derivation of the R13 and R37 monitor setpoints as discussed in Section 1.2.2
      • 2R19A setpoint calc - SC-RM-002-08, 2R19B setpoint calc - SC-RM-002*09, 2R19C setpoint calc - SC*RM-002-10, 2R19D setpoint calc
  • SC*RM-002-11.

42

Salem ODCM Rev. 8 Table 1-3 (cont'd)

Site Related Ingestion Dose Commitment Factors, Aio (mrem/hr per uCi/ml)

Nuclide Bone Liver T.Body Thyroid Kidney Lung GI-LLI Ru-103 1. 07E+2 4.60E+l 4.07E+2

1. 25E+4 Ru-105 8.89E+O 3.SlE+O l.15E+2 5.44E+3 Ru-106 l.59E+3 *2.01E+2 3.06E+3 1.03E+5 Rh-103m Rh-106 Ag-llOm 1. 56E+3 l.45E+3 8.60E+2 2.85E+3 5.91E+5 Sb-124 2.77E+2 5.23E+O l.10E+2 6.71E-l 2.15E+2 7.86E+3 Sb-125 1.77E+2 l.98E+O 4.21E+l l.SOE-1 1. 36E+2 1. 95E+3 Te-125m 2.17E+2 7.86E+l 2.91E+l 6.52E+l 8.82E+2 8.66E+2 Te-127m 5.48E+2 l.96E+2 6.68E+l 1.40E+2 2.23E+3 1. 84E+3 Te-127 8.90E+O 3.20E+O 1. 93E+O 6.60E+O 3.63E+l 7.03E+2 Te-129m 9.31E+2 3.47E+2 L47E+2 3.20E+2 3.89E+3 4.69E+3 Te-129 2.54E+O 9.55E-1 6.19E-l 1. 95E+O l.07E+l 1. 92E+O Te-131m 1.40E+2 6.85E+l 5.71E+l 1. 08E+2 6.94E+2 6.80E+3 Te-131 1. 59E+O 6.66E-1 5.03E-1 1. 31E+O 6.99E+O 2.26E-1 Te-132 2.04E+2 1.32E+2 1.24E+2 1.46E+2 1.27E+3 6. 24E+3 I-130 3.96E+l l.17E+2 4.61E+l 9.91E+3 1.82E+2 1. 01E+2 I-131 2.18E+2 3.12E+2 1.79E+2 l.02E+5 5.35E+2 8.23E+l
  • '1:-132 1. 06E+l 2.85E+l 9.96E+O 9.96E+2 4.54E+l 5.35E+O I-133 7.45E+l l.30E+2 3.95E+l l.90E+4 2.26E+2 1.16E+2 I-134 5.56E+O 1.51E+l 5.40E+O 2.62E+2 2.40E+l 1. 32E-2 I-135 2.32E+l 6.0SE+l 2.24E+l 4.01E+3 9.75E+l 6.87E+l Cs-134 6.84E+3 1.63E+4 1. 33E+4 5.27E+3 1. 75E+J'.' 2.85E+2 Cs-136 7.16E+2 2.83E+3 2.04E+3 1.57E+3 2 .16E+z};- 3.21E+2 Cs-137 .8. 77E+3 1.20E+4 7.85E+3 4.07E+3 l.35E+3 2.32E+2 Cs-138 6.07E+O 1.20E+l 5.94E+O 8.81E+O 8.70E-1 5.12E-5 Ba-139 7.SSE+O 5.59E-3 2.30E-1 5.23E-3 3.17E-3 1. 39E+l Ba-140 1.64E+3 2.06E+O l.08E+2 7.02E-1 l.18E+O 3.38E+3 Ba-141 3.SlE+O 2.SSE-3 1.29E-1 2.68E-3 1.63E-3 1. SOE-9
  • Ba-142 1.72E+O 1.77E-3 1.0SE-1 1.50E~3 l;.OOE-3 2.43E-18 La-140 1.57E+O 7.94E-1 2.lOE-1 5.83E+4 La-142 8.06E-2 3.67E-2 9.13E-3 2.68E+2 Ce-141 j.43E+o 2.32E+O 2.63E-l 1. OSE+O 8.86E+3 Ce-143 6.04E-1 4.46E+2 4.94E-2 1.97E-1 1. 67E+4 Ce-144 l.79E+2 7.47E+l 9.59E+O 4.43E+l 6.04E+4

. Pr-143 5.79E+O 2.32E+O 2.87E-l 1.34E+O 2.54E+4 Pr-144 l.90E-2 7.87E-3 9.64E-4 4.44E-3 2.73E-9 Nd-147 3.96E+O 4.58E+O 2.74E-l 2.68E+O 2.20E+4 W-187 9.16E+O 7.66E+O 2.68E+O 2.51E+3

. *~p-239 3.53E-2 3.47E-3 1. 91E-3 1.0SE-2 7.11E+2 44

Salem ODCM Rev. a Table 1-4 Bioaccumulation Factors (BFi)

(pCi/kg per pCi/liter)*

Element Saltwater Fish Saltwater Invertebrate H 9.0E-01 9.3E-Ol c 1. 8E+03 1. 4E+03 Na 6.7E-02 l.9E-Ol p 3.0E+03 3.0E+04 Cr 4.0E+02 2.0E+03 Mn 5.5E+02 4.0E+02 Fe 3.0E+03 2.0E+04 Co 1. OE+02 1. OE+03 Ni 1.0E+02 2.5E+02 Cu 6.7E+02 1. 7E+03 Zn 2.0E+03 5.0E+04 Br 1.5E-02 3.lE+OO Rb 8.3E+OO 1. 7E+Ol Sr 2.0E+OO 2.0E+Ol y 2.5E+Ol 1. OE+03 Zr 2.0E+02 8.0E+Ol Nb 3.0E+04 1. OE+02 Mo 1. OE+Ol l.OE+Ol Tc l.OE+Ol 5.0E+Ol Ru 3.0E+OO 1.0E+03 Rh 1.0E+Ol 2.0E+03 Ag 3.3E+03 3.3E+03 Sb 4.0E+Ol 5.4E+OO Te l.OE+Ol l.OE+02 I l.OE+Ol 5.0E+Ol Cs 4.0E+Ol 2. 5E+Q;l Ba 1. OE+Ol 1.0E+ot2 La 2.5E+Ol 1. OE+b3 Ce l.OE+Ol 6.0E+02 Pr 2.5E+Ol 1. OE+03 Nd 2.5E+Ol l.OE+03 w 3.0E+Ol 3.0E+Ol Np 1.0E+Ol l.OE+Ol Values in thia*table are taken from Regulatory Guide 1.109 except for phosphurus (fish) which is adapted from NUREG/CR-1336 and silver and antimony which are taken from UCRL 50564, Rev. 1, October 1972~

    • 45

tt

"!!l'

I . LE l[I I I*

I ll~ill I I 11 i I

I I

11I! i*

Ii

'\'

'_.~.

II I l-1 JlfBU I j AllO IOU-*I IO.f l

r--

-:-*- -~ . I

__ J

~1-rn~

r- -~

L - __-*-:_1_,

Salem ODCM Rev. 8

  • Table 2-1 Dose Factors for Noble Gases Total Body Ganma Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor Radionuclide Ki Li Mi Ni Cmremtyr per uCi/m3) Cmremty_r per uCi/m3) Cmrad/yr per uCi/1113) Cmrad/yr per uCi/m3)

Kr-8311 7.56E-02 1.93E+01 2.88£+02 1Cr*85m 1. 17E+03 1.46E+03 1.23E+03 1.97E+03 Kr*85 1.61E+01 1.34E+03 1.ne+o1 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03£+04 Kr-88 1.47E+04 2.37E+03 1.52£+04 2.93£+03 Kr-89 1.66E+04 1.01E+04 1.73£+04 1.06E+04 Kr*90 1.56E+04 7.29E+03 1.63£+04 7.83E+03

  • xe-131* 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133111 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe*133 2.94E+02 3.06E+02 3.53£+02 1.05£+03 Xe*13511 3.12E+03 7.11E+02 3.36E+03 7.39£+02 Xe-135 1.81E+03 1.86E+03 1.92E+03 2.46E+03 Xe-137 1.42E+03 1.22E+04 1.51E+03 1.27E+04 Xe-138 8.83E+03 4.13E+03 9.21E+03 4. 75E+03 Ar-41 8.84E+03 2.69E+03 9.30E+03 3.28E+03
  • 48

Salem ODCM Rev. a

  • Table 2-2 Parameters for Gaseous Alarm Setpoint Determinations Unit 1 Par-ter Actual Defailt lklit* ~t*

Value Value X/Q calculated 2.2E-06 sec/1l USNRC Salem Safety Evaluation. SUQ 3 VF (Plant Vent) as measured or 1.30E+OS ft 3 /min Plant Vent - normal fan curves operation (Cont Purge) 3.50E+04 Contairment Purcie AF coordinated 0.25 N/A Adninistrative allocation factor with HCGS to ensure conmined releases do not exceed release rate limit for site.

c. measured N/A uci /cm.,, Taken frOll gamna spectral analvsis of gaseous effluent Kt nuclide speci'fic N/A mretl!,}r per Values from Table 2-1 uci/

L, *nuclide specific N/A mr9{/r per Values from Table 2-1 uci/

Mi nuclide specific N/A mremt,;r per Values frOll Table 2-1 uci/

Sen 1-R41C

  • as determined 6.42E+07 Cpl per Plant Vent ***

uci/cc '

1-R16 3.6E+07 Plant Vent (Redundant) 1-R12A 2.1E+06 Contairmenti'

  • .. 1r:*

Setpoint 1-R41C calculated 1.14E+04 cpm Default alenil setpoint; more conservative values may be 1-R16 7.2E+04 (+bltg) used as deemed appropriate and desireable for ensuring 1-R12A ** 1.5E+04 (+bltg) regulatory compliance and maintaining releases ALARA *

  • Based on - calibr*tion wfth mixture of radionucl ides
    • App.licable ufng Modes 1 through 5. During Mode 6 (refueling), monitor setpoint shall be reduced to 2Jl mckgr~ in accordance with Technical Specification Table 3.3-6
      • 1R41C setpoint c*lcul*tion SC-RM-001-04
    • 49

Salem ODCM Rev. 8

  • Table 2-3 Parameters for Gaseous Alarm Setpoint Determinations Unit 2 Par-ter Actual Default Units Conmen ts Value Value X/Q calculated 2.2E-6 sec/rrf Licensing Technical S.....,,ification value VF Plant Vent as measured or 1.30E+OS ft3/min Plant Vent - normal operation fan curves Cont. Purge 3.SOE+04 Containnent Purge AF coordinated with 0.25 N/A Ac:!J!inistretive allocation factor HCGS to ensure correined releases do not exceed release rate for site.
c. measured N/A Ut;f/ar Taken fra11 ga11111111 spectral aMlvsis of gaseous effluent Ki nuclide specific N/A mremt,;r per Values from Table 2*1

. Ut;i/

L1 nuclide specific N/A mremt,;r Ut;i/

per V*lues from Table 2*1 nuclide specific N/A mremt,;r per Values frOlll Table 2-1

  • "' Ut;i/

Sen 2*R41C

  • as ctetermi ned 6.73E+07 Cf:llll per Plant Vent ***

Ut;i/CC 2*R16 3.SE+07 Pl81'1t Vent (Redl.rdant) 2-R12A 4.43E+07 Contafrrnent -**

Setpoint 2*R41C calculated 1.14E+04 CPI Default alan11 setpoints; more conservative v~s may be used 2*R16 7.2E+04 (+bkg) as deemed appr "ate and desirable for ensuring regulatory 2-R12A ** 8.60E+04 (+bkg) c~liance and for maintaining releases ALARA.

  • Based on me81'1 for c*l ibration with mixture of radionuclidea
    • Applicable m.-tna MODEi 1 through 5. During MCX>E 6 (refueling), 1110nftor setpoints shall be reduced to 2x b8ckgrCMld fn KCOrcMnc* l!lith Technical Specification 3.3-6.
      • 2R12A sepafnt calcul*tfon SC*RM-002-03, 2R41C setpoint calcul*tion SC-RM-002-07
  • so

Salem ODCM Rev. 8

  • Table 2*4 Controlling Locations, Pathways ancl Atmospheric Dispersion for Dose Calculations
  • Atmospheric Dispersion Technical Specification Location Pathway(s) Controlling ----------------------

X/Q D/Q Age Group (sec/rn3) ( 1/m2) 3.11.2.1a site bcxnlary noble gases N/A 2.2E*06 N/A (0.83 mile, N) direct exposure 3.11.2.1b site bou"tdary inhalation child 2.2E*06 N/A C0.83 *i le, N) 3.11.2.2 site bou"tdary ganma*air N/A 2.2E*06 N/A (0.83 mi le, N) beta* air 3.11.2.3 residence/dairy milk, grcx.rd infant 5.4E*08 2.1E*10 (4.9 miles, W) plane encl inhalation

  • The identified controllfnt locations, pathways encl atmospheric dispersion are frOll the Safety Evaluation Report, Suppl-.it llo. 3 for tti. .. l* Nuclear Generating Station, Unft 2 (NUREG*0517, Decetimer 1978) .
  • 51

Salem ODCM Rev. 8 Table 2-5 Pathway Dose Factors - Atmospheric Releases R(io), Inhalation Pathway Dose Factors - ADULT (mrem/yr per uCi/m3)

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 1. 26E+3 1. 26E+3 1.26E+3 1. 26E+3 1. 26E+3 1.26E+3 C-14 1. 82E+4 3.41E+3 3.41E+3 3.41E+3 3.41E+3 3.41E+3 3.41E+3 P-32 1. 32E+6 7.71E+4 8.64E+4 5.01E+4 cr-51 5.95E+l 2.28E+l 1. 44E+4 3.32E+3 1.00E+2 Mn-54 3.96E+4 9.84E+3 1.40E+6 7.74E+4 6.30E+3 Fe-55 2.46E+4 1. 70E+4 7. 21E+4 6.03E+3 3.94E+3 Fe-59 l.18E+4 2.78E+4 l.02E+6 1. 88E+5 1. 06E+4 Co-57 6.92E+2 3.70E+5 3.14E+4 6.71E+2 Co-58 1.58E+3 9.28E+5 l.06E+5 2.07E+3 Co-60 l.15E+4 5.97E+6 2.85E+5 1. 48E+4 Ni-63 4.32E+5 3.14E+4 1. 78E+5 1. 34E+4 1. 45E+4 Zn-65 3.24E+4 1. 03E+5 6.90E+4 8.64E+5 5.34E+4 4.66E+4 Rb-86 1. 35E+5 l.66E+4 5.90E+4 sr-89 3.04E+5 1.40E+6 3.SOE+S 8.72E+3 Sr-90 9.92E+7 9.60E+6 7.22E+5 6.10E+6 Y-91 4.62E+5 1.70E+6 3.85E+5 1.24E+4 Zr-95 l.07E+5 3.44E+4 5.42E+4 l.77E+6 l.50E+5 2.33E+4 Nb-95 1. 41E+4 7.82E+3 7.74E+3 5.05E+5 1. 04E+5 4.21E+3 Ru-103 1. 53E+3 5.83E+3 5.05E+5 l.lOE+S 6.58E+2 Ru-106 6.91E+4 1. 34E+5 9.36E+6 9.l~E+S 8.72E+3

.'~.

Ag-llOm 1.08E+4 1.00E+4 1.97E+4 4.63E+6 3. ()2E+5 5.94E+J Sb-124 3.12E+4 5.89E+2 7.55E+l 2.48E+6 4.06E+5 1. 24E+4 Sb-125 5.34E+4 5.95E+2 5.40E+l l.74E+6 l.01E+5 1. 26E+4 Te-125m 3.42E+J l.58E+3 l.05E+J l.24E+4 3.14E+5 7.06E+4 4.67E+2 Te-127m l.26E+4 5.77E+3 3.29E+J 4.58E+4 9.60E+5 1. 50E+5 l.57E+3 Te-129m 9.76E+3 4.67E+3 3.44E+3 3.66E+4 l.16E+6 3.83E+5 1.58E+3 I-131 2.52E+4 3.58E+4 l.19E+7 6.13E+4 6.28E+3 2.0SE+4 Cs-134 3.73E+S 8.48E+5 2.87E+5 9;76E+4 1.04E+4 7.28E+5 Cs-136 3.90E+4 1.46E+5 8.56E+4 1. 20E+4 1.17E+4 1. lOE+S Cs-137 4.78E+5 6. 21E+5 2.22E+5 7.52E+4 8.40E+3 4.28E+5 Ba-140 3.90E+4 4.90E+l 1. 67E+l 1.27E+6 2.18E+5 2.57E+3 Ce-141 1. 99E+4 *i. 35E+4 6.26E+3 3.62E+5 1.20E+S 1. 53E+3 ce-144 3.43E+6 1. 43E+6 8.48E+5 7.78E+6 8.16E+S 1.84E+5 Pr-143 9.36E+J 3.75E+3 2.16E+3 2.81E+5 2.00E+S 4.64E+2 Nd-147 5.27E+3 6.10E+3 3.56E+3 2.21E+5 1. 73E+5 3.65E+2

  • 52

. Salem ODCM Rev. 8 Table 2-5 (cont'd)

R(io), Inhalation Pathway Dose Factors - TEENAGER (mrem/yr per uCi/m3)

Nuclide Bone Liver Thyroid *Kidney Lung GI-LLI T.Body

. H-3 1. 27E+3 1. 27E+3 1. 27E+3 1.27E+3 1. 27E+3 l.27E+3 C-14 2.60E+4 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 4.87E+3 P-32 1.89E+6 1.10E+5 9.28E+4 7.16E+4 cr-51 7.50E+l 3.07E+l 2.10E+4 3.00E+3 l.35E+2 Mn-54 5.11E+4 1. 27E+4 l.98E+6 6.68E+4 8.40E+3 Fe-55 3.34E+4 2.38E+4 1.24E+5 6.39E+3 5.54E+3 Fe-59 l.59E+4 3.70E+4 1.53E+6 1. 78E+5 1. 43E+4 Co-57 6.92E+2 5.86E+5 3.14E+4 9.20E+2 Co-58 2.07E+3 1.34E+6 9.52E+4 2.78E+3 Co-60 1. 51E+4 8.72E+6 2.59E+5 1. 98E+4 Ni-63 5.80E+5 4.34E+4 3.07E+5 1.42E+4 l.98E+4 Zn-65 3.86E+4 1. 34E+5 8.64E+4 l.24E+6 4.66E+4 6.24E+4 Rb-86 1. 90E+5 1.77E+4 8.40E+4 Sr-89 4.34E+5 2.42E+6 3.71E+5 1. 25E+4 sr-90 1.0SE+S

  • 1.65E+7 7.65E+5 6.68E+6 Y-91 6.61E+5 2.* 94E+6 4.09E+5- l.77E+4 Zr-95 1.46E+S 4.58E+4 6.74E+4 2.69E+6 1. 49E+5 3.15E+4 Nb-95 1.86E+4 1. 03E+4 1. OOE+4 7.51E+5 9.68E+4 5.66E+3 Ru-103 2.10E+3 7.43E+3 7.83E+5 1.09E+5 8.96E+2 Ru-106 9.84E+4 1.90E+5 1.61E+7 9. 6C)E+5 1. 24E+4 f.,.
  • ~; ;*

Ag-llOm 1. 38E+4 1. 31E+4* 2.50E+4 6.75E+6 2.73E+5 7.99E+3 Sb-124 4. 30E+4 7.94E+2 9.76E+l 3.85E+6 3.98E+5 1. 68E+4 Sb-125 7.38E+4 8.08E+2 7.04E+l 2.74E+6 9.92E+4 1. 72E+4 Te-125m 4.88E+3 2.24E+3 1. 40E+3 5.36E+5 7.50E+4 6. 67E+2

  • Te-127m 1.80E+4 8.16E+3 4.38E+3 6.54E+4 1.66E+6 l.59E+5 2.18E+3 Te-129m 1.39E+4 6.58E+3 4.58E+3 5.19E+4 1.98E+6 4.0SE+5 2.25E+3 I-131 Cs-134 J.54E+4 5.02E+5 4.91E+4 1.13E+6
1. 46E+7 8.40E+4 - - 6.49E+3 2.64E+4 5.49E+5 3.75E+5 1.46E+5 9.76E+3 cs-136 5.15E+4 1.94E+S 1.10E+5 1.78E+4 1.09E+4 1. 37E+5 cs-137 6.70E+5 8.48E+5 3.04E+5 1.21E+5 8.48E+3 3.11E+5 Ba-140 5.47E+4 6.70E+l 2o28E+l 2.03E+6 2.29E+5 3.52E+3 Ce-141 2.84E+4 1.90E+4 8.88E+3 6.14E+S 1. 26E+5 2.17E+3 Ce-144 4.89E+6 2.02E+6 l.21E+6 1.34E+7 8.64E+5 2.62E+5 Pr-143 1. 34E+4 5.31E+3 3.09E+3 4.83E+5 2.14E+5 6.62E+2 Nd-147 7.86E+3 8.56E+3 5.02E+3 3.72E+5 1.82E+5 5.13E+2 53

Salem ODCM Rev. 8 Table 2-5 (cont'd)

R(io), Inhalation Pathway Dose Factors - CHILD (mrem/yr per uCi/m3)

Nuclide Bone Liver. Thyroid Kidney Lung GI-LLI T.Body H-3 l.12E+3 1.12E+3 1.12E+3 1.12E+3 l.12E+3 1.12E+3 C-14 3.59E+4 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3 6.73E+3*

P-32 2.60E+6 l.14E+5 4.22E+4 9.88E+4 Cr-51 8.55E+l 2.43E+l 1. 70E+4 1. 08E+3 1. 54E+2 Mn-54 4.29E+4 1.00E+4 1. 58E+6 2.29E+4 9.51E+3 Fe-55 4.74E+4 2.52E+4 1.11E+5 2.87E+3 7.77E+3 Fe-59 2.07E+4 3.34E+4 1. 27E+6 7.07E+4 1. 67E+4 Co-57 9.03E+2 5.07E+5 1.32E+4 1.07E+3 Co-58 1.77E+3 1.11E+6 3.44E+4 3.16E+3 Co-60 1. 31E+4 7.07E+6 9.62E+4 2.26E+4 Ni-63 8.21E+5 4.63E+4 2.75E+5 6.33E+3 2.80E+4 Zn-65 4.26E+4 1.13.E+5 7.14E+4 9.95E+5 1. 63E+4 7.03E+4 Rb-86 l.98E+5 7.99E+3 1.14E+5 Sr-89 5.99E+5 2.16E+6 1. 67E+5 1. 72E+4

  • sr-90 Y-91 Zr-95 Nb-95 Ru-103
1. OlE+S 9.14E+5 1.90E+5 2.35E+4 2.79E+3 4.18E+4 9.18E+3 5.96E+4 8.62E+3 7.03E+3 l.48E+7 2.63E+6 2.23E+6 6.14E+5 6.62E+5 3.43E+5 l.84E+5 6.llE+4 3.70E+4 4.48E+4 6.44E+6 2.44E+4 3.70E+4 6.55E+3 1.07E+3 Ru-106 1. 36E+5 l.84E+5 1. 43E+7 4. 2;~~+5 1. 69E+4 Ag-llOm l.69E+4 1.14E+4 2.12E+4 5.48E+6 l.OOE+5 9.14E+3 Sb-124 5.74E+4 7.40E+2 1.26E+2 3.24E+6 1.64E+5 2.00E+4 Sb-125 9.84E+4 7.59E+2 9.lOE+l 2.32E+6 4.03E+4 2.07E+4 Te-125m 6.73E+3 2.33E+3 1. 92E+3 4.77E+5 3.38E+4 9.14E+2 Te-127m 2.49E+4 8.55E+3 6.07E+3 6.36E+4 1.48E+6 7.14E+4 3.02E+3 Te-129m 1.92E+4 6.85E+3 6.33E+3 5.0JE+4 l.76E+6 l.82E+5 3.04E+3 I-131 Cs-134 4.81E+4 4.81E+4 6.51E+5 l.01E+6
1. 62E+7 7.88E+4 3.30E+5

- 2.84E+3 l.21E+5 3.85E+3 2.73E+4 2.25E+5 Cs-136 6.51E+4 1.71E+5 9.55E+4 1.45E+4 4 .18E+3 1.16E+5 cs-137 9.07E+5 8.25E+5 2.82E+5 l.04E+5 3.62E+3 1. 28E+5 Ba-140 7.40E+4 6.48E-H 2.11E+1 1.74E+6 1. 02E+5 4.33E+3 Ce-141 3.92E+4 1. 95E+4 8.55E+3 5.44E+5 5.66E+4 2.90E+3 Ce-144 6.77E+6 2.12E+6 1.17E+6 1. 20E+7 3.89E+5 3.61E+5 Pr-143 1.85E+4 5.55E+3 3.00E+3 4.33E+5 9.73E+4 9.14E+2 Nd-147 l.08E+4 8.73E+3 4.81E+3 3.28E+5 8. 21E+4 6.81E+2 54

Salem ODCM Rev. 8 Table 2-5 (cont'd)

R ( io) , Inhalation Pathway Dose Factors - INFANT (mrem/yr per uCi/m3)

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 C-14 2.65E+4 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 P-32 2.03E+6 l.12E+5 1. 61E+4 7.74E+4 cr-51 5.75E+l 1. 32E+l 1.28E+4 3.57E+2 8.95E+l Mn-54 2.53E+4 4.98E+3 l.OOE+6 7.06E+3 4.98E+3 Fe-55 1. 97E+4 l.17E+4 8.69E+4 1. 09E+3 3.33E+3 Fe=59 1. 36E+4 2.35E+4 l.02E+6 2.48E+4 9.48E+3 Co-57 6.51E+2 3.79E+5 4.86E+3 6.41E+2 Co-58 1. 22E+3 7.77E+5 l.11E+4 1. 82E+3 Co-60 8.02E+3 4.51E+6 3.19E+4 l.18E+4 Ni-63 3.39E+5 2.04E+4 2.09E+5 2.42E+3 1.16E+4 ZlT-65 1. 93E+4 6.26E+4 3.25E+4 6.47E+5 5.14E+4 3.11E+4 Rb-86 1. 90E+5 3.04E+3 8.82E+4 Sr-89 3.98E+5 2.03E+6 6.40E+4 1.14E+4 Sr-90 4.09E+7 1.12E+7 1. 31E+5 2.59E+6 Y-91 5.88E+5 2.45E+6 7.03E+4 1. 57E+4 Zr-95 1.15E+5 2.79E+4 3.11E+4 1. 75E+6 2.17E+4 2.03E+4 Nb-95 1. 57E+4 6.43E+3 4.72E+3 4.79E+S 1.27E+4 3.78E+3 Ru-103 2.02E+3 4.24E+3 5.52E+S 1. 61E+4 6.79E+2 Ru-106 8.68E+4 1.07E+S 1.16E+7 1. 6iE+5 1. 09E+4

  • .. t -~.

-'.l Ag-llOm 9.98E+3 7.22E+3 1. 09E+4 3.67E+6 3.30E+4 5.00E+3 Sb-124 3.79E+4 5.56E+2 1.01E+2 2.65E+6 5.91E+4 1. 20E+4 Sb-125 5.17E+4 4.77E+2 6.23E+l 1.64E+6 1. 47E+4 l.09E+4 Te-125m 4.76E+3 1.99E+3 1. 62E+3 4.47E+5 1. 29E+4 6.58E+2.

Te-127m 1.67E+4 6.90E+3 4.87E+3 3.75E+4 1. 31E+6 2.73E+4 2.07E+3 Te-129m 1.41E+4 6.09E+3 5.47E+3 3.18E+4 1.68E+6 6.90E+4 2.23E+3 I-131 cs;-134 3.79E+4 l.96E+5 4.44E+4 7.03E+5 l.48E+7 5.18E+4 -

1.90E+5 7.97E+4 l.06E+3

1. 33E+3
1. 96E+4 7.45E+4 Cs-136 4.83E+4 1.35E+5 5.64E+4 1.18E+4 1. 43E+3 5.29E+4 Cs-137 5.49E+5 6.12E+5 1. 72E+S 7.13E+4 1.33E+3 4.55E+4 Ba-140 5.60E+4 5.60E+l 1. 34E+l 1.60E+6 3.84E+4 2.90E+3 ce-141 2.77E+4 1.67E+4 5.25E+3 5.17E+S 2.16E+4 1. 99E+3 Ce-14.4 3.19E+6 1. 21E+6 5.38E+S 9.84E+6 1. 48E+5 1. 76E+5 Pr-143 1. 40E+4 5.24E+3 l.97E+3 4.33E+S 3.72E+4 6.99E+2 3.22E+S 3.12E+4 5.00E+2 Nd-147 7.94E+3 8.13E+3 3.15E+3 55

~ ~;:;

Salem ODCM Rev. 8 Table 2-5 (cont'd)

R(io), Grass-cow-Milk Pathway Dose Factors - ADULT (mrem/y~ per uCi/m3) for H-3 and C-14 (m2

  • mrem/yr per uCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lunq GI-LLI T.Body H-3 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 7.63E+2 C-14 3.63E+5 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 7.26E+4 P-32 1. 71E+l0 1. 06E+9 1. 92E+9 6.60E+8 Cr-51 1. 71E+4 6.30E+3 3.80E+4 7.20E+6 2.86E+4 Mn-54 8.40E+6 2.50E+6 2.57E+7 l.60E+6 Fe-55 2.51E+7 1. 73E+7 9.67E+6 9.95E+6 4.04E+6 Fe-59 2.98E+7 7.00E+7 l.95E+7 2.33E+8 2.68E+7 Co-57 1.28E+6 3.25E+7 2.13E+6 Co-58 4.72E+6 9.57E+7 1. 06E+7 co-60 1.64E+7 3.08E+8 3.62E+7 Ni-63 6.73E+9 4.66E+8 9.73E+7 2.26E+8 Zn-65 1. 37E+9 4.36E+9 2.92E+9 2.75E+9 1.97E+9 Rb-86 2.59E+9 5.11E+8 1.21E+9 sr-89 1.45E+9 2.33E+8 4.16E+7
  • sr-90 4.68E+10 1. 35E+9 l.15E+10 Y-91 8.60E+3 4.73E+6 2.30E+2 Zr-95 9.46E+2 3. 03.E+2 4.76E+2 9.62E+5 2.05E+2 Nb-95 8.25E+4 4.59E+4 4.54E+4 2.79E+8 2.47E+4 Ru-103 1.02E+3 3.89E+3 1*.19E+5 4.39E+2 RU-106 2.04E+4 3.94E+4 1. 3'2E+6

{'

2.58E+3

,_'li:

Aq-llOm 5.83E+7 5.39E+7 l*06E+8 2.2t>E+l0 3.20E+7 Sb-124 2.57E+7 4.86E+5 6.24E+4 2.00E+7 7.31E+8 1.02E+7 Sb-125 2.04E+7 2.28E+5 2.08E+4 1.58E+7 2.25E+8 4.86E+6 Te-125m 1.63E+7 5.90E+6 4.90E+6 6.63E+7 6.50E+7 2.18E+6 Te-127m 4.58E+7 1.64E+7 l.17E+7 1.86E+8 1.54E+8 S.58E+6 Te-129m 6.04B+7 2.25E+7

  • 2.08E+7 2.52E+8 3.04E+8 9.57E+6 I-131 2'.96B+8 4.24E+8 l.39E+ll 7.27E+8 1.12E+8 2.43E+S Cs-134 5.65B+9 1.34E+10 4.35E+9 1.44E+9 2.35E+8 1.lOE+lO Cs=136 2.61E+8 1.03E+9 5.74E+8 7.87E+7 l.17E+8 7.42E+8 Cs-137 7.38E+9 1.0lE+lO 3.43E+9 1.14E+9 1. 95E+8 6.61E+9 Ba-140
  • 2. 69E+7 3.38E+4 1.15E+4 1.93E+4 5.54E+7 l.76E+6 Ce-141 4.84E+3 3.27E+3 l.52E+3 1.25E+7 3.71E+2 Ce-144 3.58E+5 l.50E+5 8.87E+4 1.21E+8 1. 92E+4 Pr-143 1.59E+2 6.37E+l 3.68E+l 6.96E+S 7.88E+O
  • Nd-147 9.42E+l l.09E+2 6.37E+l 5.23E+5 6.52E+O 56

Salem ODCM Rev. 8 Table 2-5 (cont'd)

R(io), Grass-cow-Milk Pathway Dose Factors - TEENAGER (mrem/yr per uCi/m3) for H-3 and C-14 (m2

  • mrem/yr per uCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 9.94E+2 9.94E+2 9.94E+2 9.94E+2 9.94E+2 9.94E+2 C-14 6.70E+5 1. 34E+5 1. 34E+5 1. 34E+5 1. 34E+5 1. 34E+5 1. 34E+5 P-32 3.15E+10 l.95E+9 2.65E+9 1. 22E+9 Cr-51 2.78E+4 1.10E+4 7.13E+4 8.40E+6 5.00E+4 Mn-54 1.40E+7 4.17E+6 2.87E+7 2.78E+6 Fe-55 4.45E+7 3.16E+7 2.00E+7 1. 37E+7 7.36E+6 Fe-59 5. 20E+7 1.21E+8 3.82E+7 2.87E+8 4.68E+7 Co-57 2.25E+6 4.19E+7 3.76E+6 Co-58 7.95E+6 1.10E+8 1. 83E+7 Co-60 2.78E+7 3.62E+8 6.26E+7 Ni-63 l.18E+10 8.35E+8 1. 33E+8 4 *.01E+8 Zn-65 2.11E+9 7.31E+9 4.68E+9 3.10E+9 3.41E+9 Rb-86 4.73E+9 7.00E+8 2.22E+9 Sr-89 2.67E+9 3.18E+8 7.66E+7 Sr-90 6.61E+10 1.86E+9 1.63E+l0 Y-91 1.58E+4 6.48E+6 4.24E+2 Zr-95 1.65E+3 5.22E+2 7.67E+2 1. 20E+6 3.59E+2 Nb-95 1. 41E+5 7.80E+4 7.57E+4 3.34E+8 4.30E+4 Ru-103 1. 81E+3 6.40E+3 1. 52E+5 7.75E+2 Ru-106 3.75E+4 7.23E+4 ..

1.80:E+6

  • l*

4.73E+3

~g-llOm 9.63E+7 9.11E+7 1. 74E+8 2.56E+10 5.54E+7 Sb-124 4.59E+7 8.46E+5 1.04E+5 4.01E+7 9.25E+8 1.79E+7 Sb-125 3.65E+7 3.99E+5 3.49E+4 3.21E+7 2.84E+8 8.54E+6 Te-125m 3.00E+7 1.08E+7 8.39E+6 8.86E+7 4.02E+6 Te-127m 8.44E+7 2.99E+7 2.01E+7 3.42E+8 2.lOE+S 1.00E+7 Te-129m 1.11E+8 4.10E+7 3.57E+7 4.62E+8 4.15E+8 1. 75E+7 I-131 5.38E+8 7.53E+8 2.20E+ll 1.30E+9 - 1. 49E+8 4.04E+8 cs-134 9.81E+9 2.31E+10 7.34E+9 2.80E+9 2.87E+8 1.07E+l0 Cs-136 4.45E+8 1. 75E+9 9.53E+8 1.50E+8 1.41E+8 1.18E+9 Cs-137 1.34E+10 1.78E+l0 6.06E+9 2.35E+9 2.53E+8 6.20E+9 Ba-140 4.85E+7 5.95E+4 2.02E+4 4.00E+4 7.49E+7 3.13E+6 Ce-141 8.87E+3 1. 35E+4 2.79E+3 1.69E+7 6.81E+2 Ce-144 6.58E+5 2.72E+5 1. 63E+5 1. 66E+8 3.54E+4 Pr-143 1.17E+2 6.77E+l 9.61E+5 1.45E+l 2.92E+2 Nd-147 1. 81E+2 1.97E+2 l.16E+2 7.llE+S 1.18E+l 57

~

Salem ODCM Rev. 8 Table 2-5 (cont'd)

R(io), Grass-cow-Milk Pathway Dose Factors - CHILD (mrem/yr per uCi/m3) for H-3 and c-14 (m2

  • mrem/yr per uCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 1.57E+3 1. 57E+3 1.57E+3 1.57E+3 1. 57E+3 1. 57E+3 C-14 1.65E+6 3.29E+5 3.29E+5 3.29E+5 3.29E+5 3.29E+S 3.29E+S P-32 7.77E+10 3.64E+9 2.15E+9 3.00E+9 cr-51 5.66E+4 1. 55E+4 1.c53E+5 5.41E+6 1. 02E+5 Mn-54 2.09E+7 5.87E+6 1. 76E+7 5.58E+6 Fe-55 1.12E+8 5.93E+7 3.35E+7 1.10E+7 1. 84E+7 Fe-59 1. 20E+8 1. 95E+8 5.65E+7 2.0JE+S 9.71E+7 co-57 3.84E+6 3.14E+7 7.77E+6 Co-58 1.21E+7 7.08E+7 3.72E+7 Co-60 4.32E+7 2.39E+8 1. 27E+8 Ni-63 2.96E+10 1.59E+9 1. 07E+8 1. 01E+9 Zn-65 4.13E+9 1. lOE+lO 6.94E+9 1. 93E+9 6.85E+9 Rb-86 8.77E+9 5.64E+8 5.39E+9 sr-89 6.62E+9 2.56E+8 l.89E+8 Sr-90 1.12E+ll 1.51E+9 2.83E+10 Y-91 3.91E+4 5.21E+6 l.04E+3 Zr-95 3.84E+3 8.45E+2 1.21E+3 8.81E+5 7.52E+2 Nb-95 3.18E+5 1.24E+5 1.16E+5 2.29E+8 8.84E+4 Ru-103 4.29E+3 1.08E+4 1.11E+5 -1. 65E+3 Ru-106 9.24E+4 1.25E+5 1.44E+6 l.15E+4 Aq-llOm 2.09E+8 1.41E+8 2.63E+8 1.68E+10 l.13E+8 Sb-124 1.09E+8 1.41E+8 2.40E+5 6.03E+7 6.79E+8 3.81E+7 Sb-125 8.70E+7 1.41E+6 8.06E+4 4.85E+7 2.0SE+8 1. 82E+7 Te-125m 7.38E+7 2.00E+7 2.07E+7 7. l;,~E+7 9.84E+6 Te-127m 2.08E+8 5.60E+7 4.97E+7 5.93E+8 1. 68E+8 2.47E+7 Te-129m 2.72E+8 7.61E+7 8.78E+7 8.00E+8 3.32E+8 4.23E+7 I-131 1. 30E+9 l.31E+9 4.34E+ll 2.15E+9 1.17E+8 7.46E+8 Cs-134 2.26E+10 3.71E+10 1.lSE+lO 4.13E+9 2.00E+8 7. 83E+9 .

cs-136 1.00E+9 2.76E+9 l.47E+9 2.19E+8 9.70E+7 1. 79E+9 Cs-137 3.22E+l0 3.09E+10 1.0lE+lO 3.62E+9 1. 93E+8 4.SSE+9 Ba-140 l.17E+8 l.03E+5 3.34E+4 6.12E+4 5.94E+7 6.84E+6 Ce-141 2.19E+4 l.09E+4 4.78E+3 l.36E+7 l.62E+3 Ce-144 1.62E+6 5.09E+5 2.82E+S 1. 33E+8 S.66E+4 Pr-143 7.23E+2 2.17E+2 1.17E+2 7.80E+5 3.59E+l Nd-147 4.45E+2 3.60E+2 l.98E+2 5.71E+S 2.79E+l

  • 58

Salem ODCM Rev. 8 Table 2-5 (cont'd)

R(io), Grass-Cow-Milk Pathway Dose Factors - INFANT (mrem/yr per uCi/m3) for H-3 and C-14 (m2

  • mrem/yr per uCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 2.38E+3 C-14 3.23E+6 6.89E+5 6.89E+5 6.89E+5 ~.89E+5 6.89E+5 6.89E+5 P-32 1.60E+ll 9.42E+9 2.17E+9 6.21E+9.

Cr-51 1.05E+5 2.30E+4 2.05E+5 4.71E+6 1.61E+5 Mn-54 3.89E+7 8.63E+6 1. 43E+7 8.83E+6 Fe-55 1.35E+8 8.72E+7 4.27E+7 l.11E+7 2.33E+7 Fe-59 2.25E+8 3.93E+8 1.16E+8 1. 88E+8 1.55E+8 co-57 8.95E+6 3.05E+7 1. 46E+7 Co-58 2. 43E+7 . 6.05E+7 6.06E+7 Co-60 8.81E+7 2.10E+8 2.08E+8 Ni-63 J.49E+10 2.16E+9 1.07E+8 1. 21E+9 Zn-65 5.SSE+9 1.90E+10 9.23E+9 1.61E+10 8.78E+9 Rb-86 2.22E+10 5.69E+8 1.lOE+lO sr-89 1. 26E+10 2.59E+8 3.61E+8 Sr-90 1.22E+ll 1.52E+9 3.lOE+lO Y-91 7.33E+4 5.26E+6 1. 95E+3 Zr-95 6.83E+3 1.66E+3 1.79E+3 8.28E+5 l.18E+3 Nb-95 5.93E+5 2.44E+5 1.75E+5 2.06E+8 1.41E+5 Ru-103 8.69E+3 1.81E+4 1.06E+5 2.91E+3 Ru-106 1.90E+5 2.25E+5 1.44E+6 2.38E+4 Aq-llOm Sb-124 3.86E+8 2.82E+8 2.09E+8 3.08E+6 5.56E+5

- 4.03E+8

1. 31E+8 1.46E+10 1. 86E+8 6.46E+8 6.49E+7 Sb-125 1.49E+8 1.45E+6 1.87E+5 9.38E+7 1. 9*~E+8 3.07E+7 Te-125m 1. 51E+8 5. 04E+7 . 5. 07E+7 7. i:a\E+7 2.04E+7 Te-127m 4.21E+8 1.40E+8 1. 22E+8 1.04E+9 1. 70E+8 5.10E+7 Te-129m 5.59E+8 1.92E+8 2.15E+8 1.40E+9 3.34E+8 8.62E+7 I-131 2.72E+9 3.21E+9 1.05E+12 3.75E+9 1.15E+8 1. 41E+9 Cs-134 3.65E+10 6.SOE+lO 1.75E+10 7.18E+9 1.85E+8 6.87E+9 Cs-136. 1.96E+9 5.77E+9 2.30E+9 4.70E+8 8.76E+7 2.15E+9 Cs-137 5.lSE+lO 6.02E+10 1.62E+10 6.55E+9 1.88E+8 4.27E+9 Ba-140 2.41E+8 2.41E+5 5.73E+4 l.48E+5 5.92E+7 1.24E+7 Ce-141 4.33E+4 2.64E+4 8.15E+3 " - 1.37E+7 3.11E+3 Ce-144 2".33E+6 .9.52E+5 3.85E+5 1.33E+8 1. 30E+5 Pr-143 1.49E+3 '*5. 59E+2 2.08E+2 7.89E+5 7.41E+l Nd-147 8.82E+2 9.06E+2 3.49E+2 5.74E+5 5.55E+l
    • 59

Salem ODCM Rev. a Table 2-5 (cont'd)

R(io), Vegetation Pathway Dose Factors - ADULT (mrem/yr per uCi/m3) for H-3 and C-14 (m2

  • mrem/yr per uCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3* 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 2.26E+3 C-14 8.97E+5 1. 79E+5 1. 79E+5 1. 79E+5 1.79E+S 1. 79E+5 1.79E+5 P-32 1.40E+9 8.73E+7 1.58E+8 5.42E+7 Cr-51 2.79E+4 1. 03E+4 6.19E+4 1.17E+7 4.66E+4 Mn-54 3.11E+8 9.27E+7 9.54E+8 5.94E+7 Fe-55 2.09E+8 1. 45E+8 8.06E+7 8.29E+7 3.37E+7 Fe-59. 1.27E+8 2.99E+8 8.35E+7 9.96E+8 1.14E+8 Co-57 1.17E+7 2.97E+8 1. 95E+7 co-58 3.09E+7 6.26E+8 6.92E+7 Co-60 1.67E+8 3.14E+9 3.69E+8 Ni-63 1.04E+10 7.21E+8 1.50E+8 3.49E+8 Zn-65 3.17E+8 1. 01E+9 6.75E+8 6.36E+8 4.56E+8 Rb-86 2.19E+8 4.32E+7 1.02E+8 sr-s9 9.96E+9 1. 60E+9 2.86E+8 sr-90 6.05E+ll 1.75E+10 1.48E+ll Y-91 5.13E+6 2.82E+9 1.37E+5 Zr-95 1.19E+6 3.81E+5 5.97E+5 1.21E+9 2.58E+5 Nb-95 1.42E+5 7.91E+4 7.81E+4 4.80E+8 4.25E+4 Ru-103 4.80E+6 1.83E+7 5.61E+8 2.07E+6 Ru-106 1.93E+8 3.72E+8 1.25E+10 2.44E+7 Ag-llOm 1.06E+7 9.76E+6 1.92E+7 3.98E+9 5.80E+6 Sb-124 1.04E+8 1.96E+6 2.52E+5 8.08E+7 2.95E+9 4.11E+7 Sb-125 1.36E+8 1. 52E+6 1. 39E+5 l.05E+8 1.SOE+9. 3.25E+7 Te-125m 9.66E+7 3.SOE+7 2.90E+7 3.93E+8 3. 8.iE+S 1. 29E+7 Te-127m 3.49E+8 1.25E+8 8.92E+7 1.42E+9 1.11£+9 4.26E+7 Te-129m 2.55E+8 9.50E+7 8.75E+7 1.06E+9 1. 2'8E+9 4.03E+7

!~131 8.09E+7 1.16E+8 3.79E+10 1.98E+8 3.0SE+7 6.63E+7 Cs-134 4.66E+9 1.llE+lO 3.59E+9 1.19E+9 1.94E+8 9.07E+9 cs-136 4.20E+7 1.66E+8 9.24E+7 l..27E+7 l.89E+7 1.19E+8 .

Cs-137 6.36E+9 8.70E+9 2.95E+9 9.81E+8 1. 68E+8 5.70E+9 Ba-140 '1.29E+8 1.62E+5 5.49E+4 9.25E+4 2.65E+8 8.43E+6 Ce-141 1.96E+S 1.33E+5 6.17E+4 5.08E+8 1. 51E+4 Ce-144 3.29E+7 1. 38E+7 8.16E+6 1. llE+lO 1.77E+6 Pr-143 6.34E+4 2.54E+4 1.47E+4 2.78E+8 3.14E+3 Nd-147 3.34E+4 3.86E+4 2.25E+4 1.85E+8 2.31E+3

  • 60

Salem ODCM Rev. a Table 2-5 (cont'd)

R(io), Vegetation Pathway Dose Factors - TEENAGER (mrem/yr per uci/m3) for H-3 and C-14 (m2

  • mrem/yr per uCi/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 2.59E+3 C-14 1. 45E+6 2.91E+5 2.91E+5 2.91E+5 2.91E+5 2.91E+5 2.91E+5 P-32 l.61E+9 9.96E+7 l.35E+8 6.23E+7 Cr-51 3.44E+4 1. 36E+4 8.85E+4 1. 04E+7 6.20E+4 Mn-54 4.52E+8 1. J5E+8 9.27E+8 8.97E+7 Fe-55 3.25E+8 2.31E+8 1.46E+8 9.98E+7 5.38E+7 Fe-59 l.81E+8 4*. 22E+8 l.33E+8 9.98E+8 1. 63E+8 Co-57 1.79E+7 3.34E+8 3.00E+7 Co-58 4.38E+7 6.04E+8 l.OlE+S Co-60 2.49E+8 3.24E+9 5.60E+8 Ni-63 1. 61E+10 1.13E+9 1.81E+8 5.45E+8 Zn-65 4.24E+8 l.47E+9 9.41E+8 6.23E+8 6.86E+8 Rb-86 2.73E+8 4.05E+7 1. 28E+8 Sr-89 1. 51E+l0 1. 80E+9 4.33E+8 Sr-90 7.51E+ll 2.llE+lO l.85E+ll Y:-91 7.87E+6 3.23E+9 2.11E+5 Zr-95 1. 74E+6 5.49E+5 w 8.07E+5 1.27E+9 3.78E+5
  • Nb-95 Ru-103 Ru-106 Ag-llOm Sb-124 Sb-125 Te-125in
1. 92E+5 l.06E+5 6.87E+6 3.09E+8 1.52E+7 1.44E+7
1. 55E+8 2.85E+6 3.51E+5 2.14E+8 2.34E+6 2.04E+5
1. 03E+5 2.42E+7 5.97E+8 2.74E+7 1.35E+8 1.88E+8 4.55E+8 5.86E+4 5.74E+8 2.94E+6 1.48E+l0 3.90E+7 4.04E+9 8.74E+6 3.11E+9 6.03E+7 l".66E+9 5.00E+7 1.48E+8 5.34E+7 4.14E+7 4. 3*~E+8 1. 98E+7 Te-127m 5.51E+8 1.96E+8 l.31E+8 2.24E+9 1. :l'i'E+9 6.56E+7 Te-129m 3.67E+8 l.36E+8 1.18E+8 L54E+9 1. 3'sE+9 5.81E+7 I:-131 7.70E+7 1.08E+8 3.14E+10 1.85E+8 2.13E+7 5.79E+7 Cs-134 7.09E+9 1.67E+10 5.30E+9 2.02E+9 2.0SE+S 7. 74E+.9 Cs-136 4.29E+7 1.69E+8 9.19E+7 1.45E+7 1.36E+7 1.13E+8 Cs-137 1. OlE+lO 1. 35E+10 4.59E+9 1.78E+9 1.92E+8 4.69E+9 Ba-140 1.38E+8 l.69E+5 5.75E+4 1.14E+5 2.13E+8 8.91E+6 Ce-141 2.82E+5 1.88E+5 8.86E+4 5.38E+8 2.16E+4 Ce-144 Pr-143 S.27E+7 2.18E+7 7.12E+4 2.84E+4
1. 30E+7 1.65E+4

- - 1. 33E+l0 2.83E+6 2.34E+8 3.55E+3 Nd-147 3.63E+4 3.94E+4 2.32E+4 1.42E+8 2.36E+3

  • 61

Salem ODCM Rev. 8 Table 2-5 (cont'd)

R(io), Vegetation Pathway Dose Factors - CHILD (mrem/yr per uCi/m3) for H-3 and C-14 (m2

  • mrem/yr per uci/sec) for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 4.01E+3 4.01E+'3 4.01E+3 4.01E+3 4.01E+3 4.01E+3 C-14 3.50E+6 7*01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 P-32 3.37E+9 1.58E+8 9.30E+7 1. 30E+8 cr-51 6.54E+4 1.79E+4 1 .* 19E+5 6.25E+6 l.18E+5 Mn-54 6.61E+8 l.85E+8 5.55E+8 1. 76E+8 Fe-55 8.00E+8 4.24E+8 2.40E+8 7.86E+7 1. 31E+8 Fe-59 4.0lE+S 6.49E+8 1.88E+8 6. 7.6E+8 3.23E+8 Co-57 2.99E+7 2.45E+8 6.04E+7 Co-58 6.47E+7 3.77E+8 1. 98E+8 Co-60 3.78E+8 2.10E+9 l.12E+9 Ni-63 3.95E+l0 2.11E+9 l.42E+8 1. 34E+9 Zn-65 8.12E+8 2.16E+9 1. 36E+9 3.80E+8 1. 35E+9 Rb-86 sr-89 3.59E+10 4.52E+8 - 2.91E+7 2.78E+8
1. 39E+9 1. 03E+9 Sr-90 1.24E+12 1.67E+10 3.15E+ll Y-91 1. 87E+7 2.49E+9 5.01E+5 zr-95 3.90E+6 8.58E+5 1.23E+6 8.95E+8 7.64E+5 Nb-95 4.10E+5 1.59E+5 1.50E+S 2.95E+8 l.14E+5 Ru-103 l.55E+7 3.89E+7 3.99E+8 5.94E+6 Ru-106 7.45E+8 1.01E+9 1.16E+10 9.30E+7 Ag-llOm 3.22E+7 2.17E+7 4.05E+7 2.58E+9 1. 74E+7 Sb-124 3.52E+8 4.57E+6 7.78E+5 l.96E+8 2.20E+9 1. 23E+8 Sb-125 4.99E+8 3.85E+6 4.62E+5 2.78E+8 1.19E+9 1. 05E+8 Te-125m 3.51E+8 9.50E+7 9.84E+7 3. 3*~E+8 4.67E+7 Te-127m 1. 32E+9 3.56E+8 3.16E+8 3.77E+9 1. 0, 'E+9 1. 57E+8

~.!

Te-129m 8.54E+8 2.39E+8 2.75E+8 2.51E+9 1.04E+9 1. 33E+8 I-131 1.43E+8 l.44E+8 4.76E+10 2.36E+8 1. 28E+7 8.18E+7 Cs-134 1.60E+l0 2.63E+10 8.14E+9 2.92E+9 1. 42E+8 5.54E+9 Cs-136 8.06E+7 2.22E+8 l.18E+8 l.76E+7 7.79E+6 l.43E+8 cs-137 2.39E+10 2.29E+10 7.46E+9 2.68E+9 1. 43E+8 3.38E+9 Ba-140 2.77E+8 2.43E+5 7.90E+4 1. 45E+5 1.40E+8 l.62E+7 Ce-141 6.53E+5 3.26E+5 1.43E+S 4.07E+8 4.84E+4 ce-144 1.27E+8 3.98E+7 2.21E+7 - 1. 04E+l0 6.78E+6 Pr-143 1.48E+5 4.46E+4 2.41E+4 1. 60E+8 7.37E+3 Nd-147 7.16E+4 5.80E+4 3.18E+4 9.18E+7 4.49E+3

  • 62

Salem ODCM Rev. 8 Table 2-s (cont'd)

R(io), Ground Plane Pathway Dose Factors (m2

  • mrem/yr per uCi/sec)

Nuclide Any Organ H-3 C-14 P-32 Cr-Sl 4.68E+6 Mn-S4 1. 34E+9 Fe-SS Fe-S9 2.7SE+8 Co-S8 3.82E+8 Co-60 2.16E+10 Ni-63 Zn-6S 7.4SE+8 Rb-86 8.98E+6 sr-89 2.16E+4 Sr-90 Y-91 1.08E+6 Zr-9S 2.48E+8

  • Nb-9S
  • 1. 36E+8 Ru-103 1.09E+8 Ru-106 4.21E+8 Ag-llOm 3.47E+9 Te-12Sm 1.SSE+6 Te-127m Te-i29m 9.17E+4 2.00E+7

. ~ ! ~.

I-131 1. 72E+7 Cs-134 6.75E+9 cs-136 1.49E+8 cs-137 1.04E+10 Ba-140 2.0SE+7 ce-141 1. 36E+7 Ce-144 6.95E+7 Pr-143 Nd-147 8.40E+6

  • 63

APPENDIX A Evaluation of Default MPC Value for Liquid Effluents

. Appendix A Evaluation of Default MPC Value for Liquid Effluents In accordance with the requirements of Technical Specification {3.3.3.8) the radioactive liquid effluent monitors shall be operable with alarm setpoints established to ensure that the concentration of radioactive

  • material at the discharge point does not exceed the MPC value of 10 CFR 20, Appendix B, Table II, Column 2. The determination of allowable radionuclide concentration and correspondinq alarm setpoint is a function of the individual radionuclide distribution and corresponding

In order to limit the need for routinely havinq t~ reestablish the alarm setpoints as a function of changinq radionuclide di~,hibutions, a default alarm setpoint can be established. This default setpoint can be based on an evaluation of the radionuclide distribution of the liquid effluents from Salem and the effective MPC value for this distribution.

The effective MPC value for a radionuclide distribution is calculated by the equation:

  • A-2
  • Salem ODCM Rev. 8 E C1 (qamma emitters only)

MPCe = -----------------------------

Ci (qamma) Ci (non-qamma).

(A. l)

E ----------- + ~ ---------------

where:

MPCe = an effective MPC value for a mixture of radionuclide (uCi/ml)

Ci = concentration of radionuclide i in the mixture MPCi = the 10 CFR 20, Appendix. B, Table II, Column 2 MPC value for radionuclide i (uCi/ml)

The equation for determining the liquid effluent setpoints ( Section 1.2.1, equation 1.2 ) is ba~d on a multiplication of the effective MPC

  • times the monitor sensitivity. However, the *radiation monitors on the effluent lines will not detect non-gamma emitting radionuclides, such as H-3, Fe-SS, and sr-90. The derivation of the effective_MPC ( section 1.2.1, equation 1.3 ) is valid for any distribution but mu~t be modified to account for the fact that the effluent monitor will not detect the non-gammas. The above modified equation for the effective MPC provides for a default setpoint determination that accounts for the non-gamma emitting radionuclides *
  • A-3
  • Salem ODCM Rev. 8 Considering the average effective MPC value for the years 1988 through 1990, it is reasonable to select an MPCe value of 4.71E-06 uCi/ml for Unit 1 and 3.38E-06 uci/ml for Unit 2 as tYi:>ical of liquid radwaste discharges. Using these values to calculate the default Rl8 alarm setpoint value, results in a setpoint that:
1) Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations, and
2) Will provide for a liquid radwaste discharge rate (as evaluated for each batch release) that is compatible with plant operations (refer to Tables 1-1 and 1-2) .
  • A-4
  • Table A-1 Calculation of Effective MPC Salem Unit 1 Salem ODCM Rev. 8 Activity Released (Ci)

Nuclide MPC' 1988 -1989 1990 TOTAL CuCi/ml) CURIES CURIES CURIES CURIES Na-24 3E-05 1.38E-02 4.69E-04 1.69E-03 1.60E-02 Cr-51 2E-03 2.38E-02 5.25E-03 1.16£-02 4.06£-02 Mn-54 1E-04 1.01E-01 1.12E-01 1.52E-01 3.65E-01 Fe-59 5E-05 2.66E-04 1.32E-03 1.15E-03 2.73E-03 Co-57 4E-04 4.01E-03 6.11E-03 7.54E-03 1.77E-02 Co-58 9E-05 1.27E+OO 1.82E+OO 1.98E+OO 5.07E+OO Co-60 3E-05 2.77E-01 1. 78E-01 2.39E-01 6.94E-01 Zr-95 6£-05 1.23E-02 1.53E-03 4.52E-03 1.84E-02 Nb-95 1E-04 1.53E-02 3.85E-03 9. 76E-03 2.89£-02 Nb-97 9E-04 2.44E-02 7.94E-05 6.30E-03 3.0SE-02 Tc-99m 3E-03 4.74E-03 4.62E-04 8.53E-04 6.05E-03 Sr-89 3E-06 1.25E-02 1.54E-03 2.38E-03 1.64E-02 Sr-90 3E-07 2.40E-03 6.68E-04 4.66£-04 3.53E-03 Mo-99 4E-05 1.57E-03 N/0 N/D 1.57E-03 Ag-110ni 3E-05 4.96E-03 2.70E-03 8.40E-04 8.50E-03 Sn-113 BE-05 N/D N/D N/D N/D Sb-124 2E-05 6.32E-02 1.36E-02 1.94E-02 9.62E-02 Sb-125 1E-04 9.35E-02 6.53E-02 6.09£-02 2.20E-01 1-131 3E-07 5.54E-02 3.04E-02 3.53E-02 1.21E-01 1-133 1E-06 2.SOE-02 6.88E-03 8.36E-03 4.32E-02 1-134 2E-5 1.10E-02 N/D N/D 1.10E-02 1-135 4E-06 1.68E-02 1.94E-04 1.42E-04 1. 71E-02 Ce-144 1E-05 1.89E-02 1. 19E-04 1.69E-04 1.92E-02.

Cs-134 9E-06 1.31E-01 1.16E-01 1.91E-01 4.38E-01 Cs-136 6£-05 9.31E-05 9.79E-04 1.21E-03 2.28E-03 Cs-137 2E-05 1.34E-01 1.28E-01 2.02E-01 4.64E-01 Ba-140 2E-05 2.79£-04 N/D 1.10E-04 3.89E-04 La-140 2E-05 3.89£-04 2.66E-04 5.35E-04 1.19E-03 H-3 3E-03 6.34E+02 6.08E+02 3.53E+02 1.59E+03 Fe-55 SE-04 5.40E-01 1. 75E-01 1.61E-01 8.76E-01 ~

W-187 6£-05 1.25E-02 N/D N/D 1.25E-02 . 'ii.

Zn-65 1E-04 5.49E-04 3.62E-04 7.75E-03 8.66E-03 ~:

Zr-97 2E-05 1.37E-02 N/D N/D 1.37E-02 Total C. G1111111a 2.33E+OO 2.49E+OO 2.94E+OO 7.77E+OO Total C. Non-ganma 6.35E+02 6.08E+02 3.53E+02 1.60E+03 MPC, CUCi/111l) 4.71E-06 6.88E-06 9.45E-06 MPC value for 111r..tricted are* from 10 CFR 20, Appendix B, Table II, CotU111 2. *

    • N/D - not detected
  • A-5

d Salem ODCM Rev. 8 Table A-2 Calculation of Effective MPC Salem Unit 2 Activity Released (Ci)

MPC 1988 1989 1990 TOTAL Nuclide CuCi/ml) CURIES CURIES CURIES CURIES Na-24 3E-05

1. 04E-02 8.0SE-04 2.28E-03 1.3SE-02 Cr-51 2E-03 3. 1'7E-03 1.57E-02 1.48E-02 3.37E-02 Mn-54 1E-04 1. 74E-01 1.19E-01 1.52E-01 4.4SE-01 Fe-59 5E-05 2.93E-05 3.00E-03 1.09E-03 4. 12E-03 Co-57 4E-04 4.55E-03 6.70E-03 7.92E-03 1.92E-02 Co-58 9E-05 1.32E+OO 2.02E+OO 2.01E+OO S.35E+OO Co-60 3E-05 2.97E-01 2.08E-01 2.36E-01 7.41E-01

. Zr-95 6E-05 3.1SE-03 3.39E-03 5.22E-03 1.18E-02 Nb-95 1E-04 6.55E-03 7.41E-03 1.03E-02 2.42E-02 Nb-97 9E*04 6.92E-03 2.54E-04 5.32E-04 7.71E*03 Tc-99m 3E-03 3.28E-03 6.64E-04 8.66E-04 4.81E*03 Sr-89 3E-06 1.69E*02 1.52E*03 2.28E-03 2.07E-02 Sr-90 3E-07 4.11E-03 6.45E-04 4.73E*04 5.23E-03 Mo-99 4E*OS 1.19E*04 N/D N/D 1.19E-04 Ag*110m 3E*05 1.04E*02 6.41E*03 2.56E-03 1.94E*02 Sn-113 8E*OS N/D N/D N/D N/D Sb-124 2E*OS 5.47E-02 1.89E*02 2.22E*02 9.58E-02 Sb-125 1E*04 9.22E*02 8.0SE-02 7.40E*02 2.47E-01 1*131 3E*07 1.3SE*01 3. 79E-02 3.83E*02 2.11E*01 1-133 1E*06 8.83E*02 8.64E*03 1.07E-02 1.0SE-01 I *134 2E*05 3.49E*02 N/D N/D 3.49£-02 I-135 4E*06 1.90E*02 5.17E-04 7.09E-04 2.02E*02 Ce-144 1E*OS 2.24E*03 6.05E*04 7.67E-05 2.92E*03 Cs-134 9E-06 9.53E*02 1.43E*01 1.86E-01 4.24E*01 Cs-136 6E*OS 2.20E*03 1.39E*03 1.31E*03 4.90E-03 Cs-137 2E*OS 1.09E*01 1.55E*01 1.95E*01 4.59E-01 Ba-140 2E*OS 1.57E*03 N/D N/D 1.57E-03 La-140 2E*OS 1.03E*03 5.19E*04 6.23E*04 2.17E-03 H*3 3E-03 3.68E+02 5.02E+02 3.03E+02 1.17E+03 Fe-55 SE-04 4.69E-01 1.84E*01 2.09E-01 8.62E*01 W*187 6E-05 6.37E-04 N/D N/D 6.37E-04 ....;

Zn-65 1E-04 11/D 1.41E*04 1.06E*02 1.07E-02 . t:.

Total Ci G&1111111 2.48E+OO 2.84E+OO 2.98E+OO 8.30E+OO Total Ci Non-gamna 3.68E+02 5.02E+02 3.03E+02 1.17E+03 MPC0 (uCi /ml) 3.38E-06 7.85E*06 9.71E-06 MPC value for ...,restricted ere* from 10 CFR 20, Appendix B, Table II, Colllll'I 2.

    • N/D - not detected
    • A-6
  • Salem ODCM Rev. 8 APPENDIX B Technical Basis for Effective Dose Factors Liquid Radioactive Effluent
  • B-1
  • APPENDIX B Salem ODCM Technical Basis for Effective Dose Factors -

Liquid Effluent Releases Rev. 8 The radioactive liquid effluents for the years 1982 through 1989 were evaluated to determine the dose contribution of the radionuclide distribution. This analysis was performed to evaluate the use of a limited dose analysis for determining environmental doses, providing a simplified method of determining compliance with the dose limits of Technical Specification 3.11.1.2. For the radionuclide distribution of effluents from Salem, the controlling organ is the GI-LLI. For the last three years the calculated GI-LLI dose is predominately a function of the Fe-55, co-58, Co-60 and

  • Nb-95 releases. The radionuclides, Co-58 and cs-134 contribute the large majority of the calculated total body dose. The results of the evaluation for 1989, 1988, and 1987 are presented in Table B-1 and Table B-2.

For purposes of simplifying the details of the dose calculational process, it is conservative to identify a controlling, dose significant radionuclide and limit the calculation process to the use of the dose conversion factor for this nuclide. Multiplication of the total release (i.e., cumulative activity for all radi'onuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.

For the evaluation of the maximum organ dose, it is conservative to

  • B-2
  • Salem ODCM Rev. 8 use the Nb-95 dose conversion factor (1.51 E+06 mrem/hr per uCi/ml, GI-LLI). By this approach, the maximum organ dose will be overestimated since this nuclide has the highest organ ddse factor of all the radionuclides evaluated. For the total body calculation, the Fe-59 dose factor (7.27 E+04 mrem/hr per uci/ml, total body) is the highest among the identified dominant nuclides. For evaluating compliance with the dose limits of Technical Specification 3.11.1.2, the following simplified equations may be used:

Total Body 1.67E-02 '* VOL Dtb = cw

  • A Fe-59,TB * (B .1)
  • where:

Dlb A Fe-59,TB

=

=

dose to the total body {mrem) 7.27E+04, total body ingestion dose conversion factor for Fe-59 {mrem/hr per uci/ml) ,:\

VOL = volume of liquid effluent released {gal}\'*

Ci = total concentration of all radionuclides (uci/ml) cw = average circulating water discharge rate during release period(gal/min)

1. 67E-02 = conversion factor (hr/min)

Substituting the value for the Fe-59 total body dose conversion factor, the equation simplifies to:

.1.21B+03

  • VOL cw * (B.2)
  • B-3
  • Maximum organ Salem ODCM Rev. 8 l.67E-02
  • VOL *A Nb-95,GI-LLI Dmu = --------------------------- * (B.3) where:

Dmax = maximum organ dose (mrem)

A Nb-95,GI-LLI = 1.51E+06, Gi-LLI ingestion dose conversion factor for Nb-95 (mrem/hr per uCi/ml)

Substituting the value for A Nb-95,GI-LLI the equation simplifies to:

2.52E+04

  • VOL Dmax - * (B. 4) cw Tritium is not included in the limited analysis dose assessment for liquid releases, be~ause the potential dose resulting fro~ normal

.. \;.

reactor releases is relatively negligible. The average annual tritium release from each Salem Unit is approximately 350 curies.

The calculated total body dose from such a release is 2.4E-03 mrem/yr via the fish and invertebrate ingestion pathways. This amounts to 0.08% of the design objective dose of 3 mrem/yr.

Furthermore, the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste system operation .

  • B-4

Salem ODCM Rev. 8 Table B-1 Adult Dose Contributions Fi sh. and Invertebrate Pathways Unit 1 1989 1988 1987 Radio* RELEASE TBOOY GI-LLI LIVER RELEASE TBODY GI-LLI LIVER RELEASE TBOOY Gl-LLI LIVER nuclide (Ci) Dose Dose Dose (Ci) Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac MN-54 1.12E-01 0.06.

0.03 0.11 1.01E-01 0.01 0.03 0.03 1.05E-01 a.a1 a.a5 a.a4 FE-55 3.98E-a2 a.as a.a2 a.19 5.44E-01 a.43 a.11 0.76 2.35E-01 a. 16 a.11 a.43 FE-59 1.32E-a3 a.a2 a.a2 a.a3 2.66E-04 * *

  • N/D * *
  • C0-58 1.82E+OO a.39 a.56 a.16 1.27E+aO a. 11 a.24 a.a3 1.54E+OO a.17 a.42 a.as C0-60 1. 78E-01 a.11 0.15 a.a4 2..77E-01 a. 10 0.14 0.02 4.21E-01 0. 13 a.31 a.a4 ZN-65 3.62E-a4 a.a1
  • a.a2 5.49E-a4 a.a1
  • a.01 N/D * *
  • NB-95 3.8SE-03
  • a. 15
  • 1.53E-02
  • a.36
  • 2.44E-a3
  • a.a8
  • AG-110M 2.7aE-a3
  • a.04
  • 4.96E-03
  • o.os
  • 2.36E-03
  • a.o3
  • CS* 134 1.16E-01 0.24
  • a.25 1.31E-01 0.17
  • 0.08 3.11E-01 0.34
  • a.21 1.34E-a1 a.10
  • a.a6 3.a1E-01 a.19
  • a.19 Total 2.4aE+aO 2.48E+OO 2.92E+OO Table i-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1989 1988 1987 ISOTOPE RELEASE TBOOY GI-LLI LIVER RELEASE TBOOY Gl*LLI LIVER RELEASE TBOOY GI -LLI LIVER (Ci) Dose Dose Dose (Ci) Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac.

MN-54 1.19E-01 0.02 0.05 o.a9 1.74E-01 0.03 0.07 0.06 1.20E-01 o.a1 a.04 a.a2 FE-55 4.61E-02 0.05 0.02 0.18 4.69E-01 0.42 0.16 0.75 8.74E*01 0.39 0.26 a.72 FE-59 3.00E-03 0.03 0.04 0.06 2.93E-05 * *

  • N/D * * **

C0-58 2.02E+OO 0.37 0.47 a.14 1.32E+OO 0.19 0.29 0.04 1. 71E+OO -0.12 0.31 a.a2 C0-60 2.08E-01 . 0.11 0.13 0.04 2.97E-01 a.12 0.18 0.02 4.23E-01 .. 'i}.a9 0.21 o.a2 ZN-6S 1.41E-04 *

  • a.01 N/D * *
  • N/D .~ . * *
  • NB-95 7.41E-03
  • a.22
  • 6.55E-03
  • 0.18
  • 7.92E-03
  • o. 18 AG-110M 6.41E-03
  • 0.07
  • 1.a4E-02
  • 0.11
  • N/D * * **

CS-134 1.43E-01 0.25

  • 0.26 9.53E-02 0. 14
  • 0.07 3.49E-01 0.25
  • 0.21 1.09E-01. 0.09
  • 0.06 3.33E-01 o. 14
  • o.a9 Total 2.71E+OO 2.48E+OO 3.82E+OO
  • less than 0.01 N/D = not detected
  • B-5

Salem OOCM Rev. 8 APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent

~.'*

  • c-1

Salem OOCM Rev. 8

  • overview APPENDIX C Technical Bases for Effective Dose Factors -

Gaseous Radioactive Effluents The evaluation of doses due to releases of radioactive material to the atmosphere can be simplified by the use of effective dose transfer factors instead of using dose* factors which are radionuclide specific. These effective factors, which can be based on typical radionuclide distributions of releases, can be applied to the total radioactivity released to approximate the dose in the environment (i.e., instead of having to perform individual radionuclide dose analyses only a single multiplication (Kcffi Meff or Neff) times the total quantity of radioactive material released would be needed)

  • This approach provides a reasonable estimate of the actual dose while eliminating the need for a detailed calculational technique.

~ :

Determination of Effective Dose Factors Effective dose transfer factors are calculated by the following equations:

- * (C.1) where:

= the effective total body dose factor due to gamma emissions from all noble gases released

= the total body dose factor due to gamma emissions from each noble gas radionuclide i released

= the fractional abundance of noble gas radionuclide i relative to the total noble gas activity

    • c-2

Salem OOCM Rev. 8

  • (C. 2) where:

(L + 1.1 M)eff =*the effective skin dose factor due to beta and gamma emissions from all noble gases released

(~ + 1.1 ~) = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released (C. 3) where:

MctJ = the effective air dose factor due to gamma emissions from all noble gases released

= the air dose factor due to gamma emissions from each noble gas radionuclide i released (C. 4) where:

= the effective air dose factor due to beta emissions from all noble gases released

= the air dose factor due to beta emissi~ns f~pm each noble gas radionuclide i released Normally, it would be expected that past radioactive effluent data would be used for the determination of the effective dose factors.

However, the noble gas releases from Salem have been maintained to such negligible quantities that the inherent variability in the data makes any meaningful evaluations difficult. For the past years, the total noble* gas releases have been limited to* 2,000 Ci for 1984, C-3

Salem ODCM Rev. 8

  • 2,800 Ci for 1985, 2,700 Ci for 1986, 1700 Ci for 1988, and 1500 Ci for 1989. Therefore, in order to provids a reasonable basis for the derivation of the effective noble gas dose factors, the primary coolant source term from ANSI N237-1976/ANS-18.1, "Source Term Specifications," has been used as representing a typical distribution. The effective dose factors as derived are presented in Table C-1.

Application To provide an additional degree of conservatism, a factor of 0.50 is introduced into the dose calculational process when the effective dose transfer factor is used. This conservat~sm provides additional assurance that the evaluation of doses by the use of a single effective factor will not significantly underestimate any actual doses in the environment. For evaluating compliance with:the dose limits of Technical Specification 3.11.2.2, the following simplified equations may be used:

3.17B-08 Di: = --------

o.so

  • X/Q
  • M.:tr * :E Qi (C. 5) and 3.17E-08

~ = --------

o.so

  • X/Q
  • Neff
  • :E Qi (C. 6)
  • C-4

Salem ODCM Rev. 8

  • where:

Dg = air dose due to gamma emissions for the cumulative release of all noble gases (mrad)

Db air dose due to beta emissions for the cumulative releas.e of all noble gases (mrad)

X/Q = atmospheric dispersion to the controlling site boundary (sec/m3)

Melf = 5.3E+02, effective gamma-air dose factor (mrad/yr per uCi/m3)

Neff = 1.1E+03, effective beta-air dose factor (mrad/yr per uCi/m3)

Qi = cumulative release for all noble gas radionuclides (uCi}

3.17E-08 = conversion factor (yr/sec}

0.50 = conservatism factor to account for the variability in the effluent data Combining the constant~, the dose calculational equations simplify to:

DI = 3.SE-05 and

  • X/Q
  • E Qi (C. 7)

= 7.0E-05

  • X/Q
  • E Qi *~JC. 8)

The effective dose factors are used on a very limited basis for the purpose of facilitating the timely assessment of radioactive effluent releases, particularly during periods of computer malfunction where a detailed dose assessment may be unavailable .

c-s

Salem ODCM Rev. 3

  • Table C-1 Effective Dose Factors Noble Gases - Total Body and Skin Total Body Effective SkinEffective Radionuclide Dose Factor Dose Factor K.tr ( L+ 1. 1 M) etr (mrem/yr per uCi/m3 ) (mrem/yr per uCi/m3 )

Kr-85 0.01 l.4E+Ol Kr-88 0.01 l.5E+02 l.9E+02 Xe-133m 0.01 2.5E+OO l.4E+Ol Xe-133 0.95 3.0E+02 6.6E+02 Xe-135 0.02 3.6E+Ol 7.9E+Ol Total 4.8E+02 9.6E+02 Noble Gases - Air Gamma Air Effective Beta Air Effective Radionuclide Dose Factor Dose Factor M.tr Neff (mrad/yr per uCi/~) (mrad/yr per -uci/m3 )

Kr-85 0.01 2.0E+Ol Kr-88 0.01 l.5E+02 2.9E+Ol Xe-133m 0.01 3.3E+OO 1. SE+Ol Xe-133 0.95 3 . .4E+02 1. OE+03 Xe-135 0.02 3.8E+Ol 4 :;9E+Ol Total 5.3E+02 1. 1E+03

  • Based on Noble gas distribution from ANSI N237-1976/ANSI-18.l, "Source Term Specifications."

C-6

Salem ODCM Rev. 8 APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1

Salem ODCM Rev. 8

  • APPENDIX D Technical Basis for* Effective Dose Parameter Gaseous Radioactive Effluent Releases The pathway dose factors for the controlling infant age group were evaluated to determine the controlling pathway, organ and radionuclide. This analysis was performed to provide a simplified method for determining compliance with Technical Specification 3.11.2.3 For the infant age group, the controlling pathway is the grass-milk-cow (g/m/c) pathway. An infant receives a greater radiation dose from the g/m/c pathway than any other pathway. Of this g/m/c pathway, the maximum exposed organ including the total
  • body, is the thyroid, and the highest dose contributor is radionuclide I-131.

in Table D-1.

The results for this evaluation are presented For purposes of simplifying the details of the dose calculation process, it is conservative to identify a controlling, dose significant organ and radionuclide and limit the calculation process to the use of the dose conversion factor for the organ and radionuclide. Multiplication of the total release (i.e. cumulative activity for all radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative .

  • D-2

Salem ODCM Rev. 3

  • For the evaluation of the dose commitment via a controlling pathway and age group, it is conservative to use the infant, g/m/c, thyroid, I-131 pathway dose factor (1.05E12 m2 mrem/yr per uCi/sec). By this approach, the maximum dose commitment will be overestimated since I-131 has the highest pathway dose factor of all radionuclides evaluated.

For evaluating compliance with the dose limits of Technical Specification 3.11.2.3, the following simplified equation may be used:

  • where:

Dmax w

Dmu:

=

=

= 3.17E-8

  • W
  • RI-131 maximum organ.dose (mrem)
  • E Qi atmospheric dispersion parameters to the
  • controlling location(s) as identified~~n Table 3.2-4. ,

X/Q = atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways (sec/m3 )

D/Q = atmospheric deposition for vegetation, milk and ground plane exposure pathways (m.2 )

Qi = cumulative release over the period of interest for radioiodines and particulates 3.17E-8 = conversion factor (yr/sec)

RI-131 = I-131 dose parameter for the thyroid for the identified controlling pathway

= 1.05E12 (m2 mrem/yr per uCi/sec), infant thyroid dose parameter with the cow-milk=grass pathway controlling The ground plane exposure and inhalation pathways need not be

., considered when the above simplified calculation method is used because for the overall negligible contribution of these pathways to D-3

Salem ODCM Rev. 3

  • the total thyroid dose. It is recognized that for some particulate radioiodines (e.g., Co-60 and Cs-137), the ground exposure pathway may represent a higher dose contribution than either the vegetation or milk pathway. However, use of the I-131 thyroid dose parameter for all radionuclides will ma~imize the organ dose calculation, especially considering that no other radionuclide has a higher dose parameter for any organ via any pathway than I-131 for the thyroid via the milk pathway (see Table D-1)
  • The location of exposur~ pathways and the maximum organ so calculation may be based on the available pathways in the
  • surrounding environment of Salem as identified by the annual land-use census (Technical Specification.3.12.2). Otherwise, the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4 *
  • D-4

Salem OOCM Rev. 8

  • Table D-1 Infant Dose Contributions Fraction of Total Organ and Body Dose PATHWAYS Target Organs Grass-Cow-Milk Ground Plane Total Body 0.02 0.15 Liver 0.23 0.14 Thyroid 0.59 0.15 Kidney 0.02 0.15 Lung 0.01 0.02 GI-LLI 0.02 0.15 Fraction of Dose Contribution .Q:l Pathway Pathway Grass-Cow-Milk 0.92 Ground Plane 0.08 Inhalation *
  • D-5

Salem ODCM Rev. 8 APPENDIX E Radiological Environmental Monitoring Program Sample Type, Location and Analysis

\:

  • E-1

Salem ODCM Rev. 3

  • APPENDIX E SAMPLE DESIGNATION Samples are identified by a three part code. The first two letters are the power station identification code, in this case "SA". The next three letters are for the media sampled.

AIO = Air Iodine IDM = Immersion Dose (TLD)

APT = Air Particulates MLK = Milk ECH = Hard Shell Blue Crab PWR = Potable Water (Raw)

ESF = Edible Fish PWT = Potable Water (Treated)

ESS = Sediment RWA = Rain Water (Precipitation)

FPB = Beef SWA = Surf ace Water FPL = Green Leafy Vegetables VGT = Fodder Crops (Various)

FPV = Vegetable (Various) WWA = Well Water GAM = Game The last four symbols are a location code based on direction and distance from the site. Of these, the first two represent each of the sixteen angular sectors of 22.5 degrees centered about the reactor site. Sector one is divided evenly by the north axis and

  • other sectors are numbered in a clockwise direction; i.e., 2=NNE, 3=NE, 4=ENG, etc. The next digit is a letter which represents the radical distance from the plant:

s =

A =

on.:..site location 0-1 miles off-site E

F

=

=

4-5 miles off-site 5-10 miles off-site B = 1-2 miles off-site G = 10-20 miles off-si~e c = 2-3 miles off-site H = > 20 miles off-sit.

D = 3-4 miles off-site The last number is the station numerical designation within each sector and zone; e.g., 1,2,3, ... For example; the designation SA-WWA-501 would indicate a sample in the SGS program (SA),

consisting of well water (WWA), which had been collected in sector number 5, centered at 90' (due east) with respect to the.reactor site at a radical distance of 3 to 4 miles off-site; (therefore, radial distance D). The number 1 indicated that this is sampling station #1 in that particular sector .

  • E-2

Salem ODCM Rev. 3

  • SAMPLING LOCATIONS All sampling locations and specific information about the individual locations are given in Table E. Maps E-1 and E-2 show the locations of sampling stations with respect to the site.

TABLE E-1 STATION CODE STATION LOCATION SAMPLE TYPES 2S2 0.4 mi. NNE of vent IDM 3S3 700 ft. NNE of vent; fresh water WWA holding tank 5Sl 1.0 mi. E of vent; site access road AIO, APT, IDM 6S2 0.2 mi. ESE of vent; observation IDM building 7SI 0.12 mi. SE of vent; station personnel IDM gate

  • lOSl llSl llAl 0.14 mi. SSW of vent; site shoreline.

0.09 mi. SW of vent; site shoreline 0.2 mi. W of vent; outfall area IDM IDM ECH, ESF, ESS, SWA **

'te.

15Al 0.3 mi. NW of vent; cooling tower ESS blowdown discharge line 16Al 0.7 mi. NNW of vent; south storm drain ESS discharge line 12Cl. 2.5 mi. WSW of vent; west bank of ECH, ESF, ESS Delaware River SWA 4D2 3.7 mi. ENE of vent; Alloway Creek IDM Neck Road 5Dl 3.5 mi. E of vent; local farm AIO, APT, IDM, WWA lODl 3.9 mi. SSW of vent; Taylor's Bridge IDM Spur

  • E-3

Salem ODCM ~ev. 3

  • TABLE E-1 (Cont'd)

STATION CODE STATION LOCATION SAMPLE TYPES llDl 3.5 mi. SW of vent GAM 14Dl 3.4 mi. WNW of vent; Bay View, Delaware IDM 2El 4.4 mi. NNE of vent; local farm IDM 3El 4.1 mi. NE of vent; local FPB, FPV, GAM, IDM, VGT, WWA 3F2 5.7 mi. NE of vent; local farm FPV 7El 4.5 mi. SE of vent; 1 mi. W of Mad. ESF, ESS, SWA Horse Creek 9El 5.0 mi. SW of vent IDM 11E2 5.0 mi. SW of vent IDM 12El 4.4 mi. WSW of vent; Thomas Landing IDM 13El 4.2 mi. w of vent; Diehl House Lab IDM 13E3 4.9 mi. w of vent; local VGT '*

1*

16El 4.1 mi. NNW of vent; Port Penn AIO, APT, IDM lFl 5.8 mi. N of vent; Fort Elf sborg AIO, APT, IDM 1F2 7.1 mi. N of vent; midpoint of SWA Delaware 1F3 5. 9' mi. N of vent; local farm FPL, FPV 2F2 8.7:mi. NNE of vent; Salem Substation AIO, APT, IDM, RWA 2F3 8.0 mi. NNE of vent; local farm FPV 2F4 6.3 mi. NNE of vent; local FPV 2F5 7.5 mi. NNE of vent; Salem High School IDM

    • E-4

Salem ODCM Rev. 8 TABLE E-1 (Cont'd)

  • STATION CODE 2F6 STATION LOCATION 7.3 mi. NNE of vent; Southern Training center SAMPLE TYPES IDM 2F7 5.7 mi. NNE of vent; local farm MLK, VGT 3F2 5.1 mi. NE of vent; Hancocks Bridge IDM Municipal Building 3F3 8.6 mi. NE of vent; Quinton Township IDM School 5Fl 6.5 mi. E of vent FPV,IDM 5F2 7.0 mi. E of vent; local farm VGT 6Fl 6.4 mi. ESE of vent; Stow Neck Road IDM 7F~ 9.1 mi. SE of vent; Bayside, NJ IDM 10F2 5.8 mi. SSW of vent IDM llFl 6.2 mi. SW of vent; Taylor's Bridge IDM Delaware 11F3 5.3 mi. SW of vent; Townsend, DE MLK, VGT 12Fl 9.4 mi. WSW of vent; Townsend Elem. IDM' School 13F2 6.5 mi. W of vent; Odessa, DE 13F3 9.3 mi. W of vent; Redding Middle IDM School, Middletown, DE 13F4 9.8 mi. W of vent; Middletown, DE IDM 14Fl 5.5 mi. WNW of vent; local farm VGT 14F2 6.6 mi. WNW of vent; Boyds Corner IDM 14F3 5.4 mi. WNW of vent; local farm FPV 14F4 7.6 mi. WNW of vent; local farm MLK 15F3 5.4 mi. NW of vent IDM
  • E-5 *I I

I i

.Salem ODCM Rev. 8

'-* STATION CODE TABLB B-1 (Cont'd)

STATION LOCATION SAMPLE TYPES 16Fl 6.9 mi. NNW of vent; C&D Canal ESS, SWA 16F2 8.1 mi. NNW of vent; Delaware City IDM Public School lGl 10.3 mi. N of vent; local farm FPV 1G3 19 mi. N of vent; Wilmington, DE IDM 2Gl 12 mi. NNE of vent; Mannington FPV Township, NJ 3Gl 17 mi. NE of vent; local farm IDM, MLK, VGT lOGl 12 mi. SSW of vent; Smyrna, DE IDM 16Gl 15 mi. NNW of vent; Greater Wilminqton IDM Airport

  • 3Hl 3H3 3H5 32 mi. NE of vent; National Park, NJ 110 mi. NE of vent; Research and Testing 25 mi. NE of vent; local farm IDM AIO, APT, IDM FPL,. FPV

Salem_ OOCM Rev. 8

  • Sample SAMPLES COLLECTION AND ANALYSIS Collection Method Analysis Air Particulate Continuous low volume Gross Beta analysis air sampler. Sample on each weekly collected every week sample. Gamma along with the filter spectrometry shall change. be performed if gross beta exceeds 10 times the yearly mean of the control station value. As well one sample is analyzed > 24 hrs after sampling to allow for radon and thorium daughter decay. Gamma isotopic analysis on quarterly composites.

Air Iodine A TEDA impregnated Iodine 131 analysis charcoal cartridge is are performed on connected to air each weekly sample.

particulated air sampler and is collected weekly

  • at filter change.

Crab and Fish Two batch samples are Gamma *isotopj_..Ct:

sealed in a plastic analysis of edible bag or jar and frozen portion on collection.

semi-annually or when in season.

Sediment A sediment sample is Gamma isotopic taken semi-annually. analysis semi-annually.

Direct 2 TLD's will be. Gamma dose quarterly.

collected from each location quarterly.

-* E-7

Salem ODCM Rev. 8

  • Sample SAMPLE COLLECTION AND ANALYSIS (Cont'd)

Collection Method Analysis Milk Sample of fresh milk Gamma isotopic is collected for each analysis and I-131 farm semi-monthly when analysis on each cows are in pasture,

  • sample on collection.

monthly at other times.

Water (Rain, Sample to be collected Gamma isotopic Potable, monthly providinq winter monthly H-3 onsurf ace) icinq conditions allow. quarterly surface sample, monthly on qround water sample .

....~.

    • E-8
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