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{{#Wiki_filter:Salem Generating Station Revision # 6 of the ODCM Revision Date: 4/11/90 Unit No. 50-272 Unit 2 Docket No. 50-311 Operating License No. DPR-70 Operating License No. DPR-75 ** 1 *. J I . ,
{{#Wiki_filter:Salem Generating Station Revision # 6 of the ODCM Revision Date: 4/11/90 Unit ~ Dock~t No. 50-272 Unit 2 Docket No. 50-311 Operating License No. DPR-70 Operating License No. DPR-75
SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Revision 6 03/28/90 Approval SORC Date:
                                  *.
* Salem ODCM Rev. 6 03/28/90 SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents Introduction . . . . . . . . . . . . .
                                  ** 1 J
1 1. 0 2.0 3.0 Liquid Effluents 1.1 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 ....... 5 1.3 Liquid Effluent Concentration Limits -10 CFR 20 7 1.4 Liquid Effluent Dose Calculations  
I
-10 CFR 50 . . . 8 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  
PS~G
..........
 
13 Gaseous Effluents 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  
SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Revision 6 03/28/90 Approval SORC Chairman:~;          Date:
-10 CFR 20 . . . . 2.3.1 Site Boundary Dose Rate -Noble Gases 2.3.2 Site Boundary Dose Rate -. 20
 
* 2 0 Radioiodine and Particulates . * . . . * . . 21 2.4 Noble Gas Effluent Dose Calculations  
Salem ODCM Rev. 6 03/28/90
-10 CFR 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 NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Table of Contents Introduction . .   . . .                       ........               1
-10 CFR 50 * * * * * * * * * * * * * * * * *
: 1. 0   Liquid Effluents 1.1 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         .......         5 1.3 Liquid Effluent Concentration Limits - 10 CFR 20           7 1.4 Liquid Effluent Dose Calculations - 10 CFR 50 . . .         8 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 2.0    Gaseous Effluents 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
* 2 6 2.5.1 UNRESTRICTED AREA Dose -Radioiodine and Particulates . . . . 27 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates . * . * * . * . . . 27 -2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations . . * . 2.7 Gaseous Effluent Dose Projection Special Dose Analyses . 28 . . 32 3.1 Doses Due To Activities Inside the SITE BOUNDARY .. 33 3.2 Doses to MEMBERS OF THE PUBLIC -40 CFR 190 .... 33 3.2.1 Effluent Dose Calculations  
* 20 2.3.2 Site Boundary Dose Rate -
....... 35 3.2.2 Direct Exposure Determination  
Radioiodine and Particulates . * . . . * . .       21 2.4 Noble Gas Effluent Dose Calculations - 10 CFR 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
....... 35
              - 10 CFR 50   * * * * * * * * * * * * * * *     * *
.-Salem ODCM Rev. 6 03/28/90 Table of Contents -Continued 4.0 Radiological Environmental Monitoring Program .
* 26 2.5.1   UNRESTRICTED AREA Dose -
* 36 4.1 Sampling Program ............. .
Radioiodine and Particulates             . . . . 27 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates . * . * *         . * . .       . 27
* 36 4.2 Interlaboratory Comparison Program ... . *
        -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 . . . .         33 3.2.1 Effluent Dose Calculations           .......         35 3.2.2 Direct Exposure Determination       .......         35
* 3 7 Tables 1-1. Parameters for Liquid Alarm Setpoint Determination 1-2 1-3 1-4 2-1 2-2 2-3 2-4 2-5 A-1 A-2 B-1 B-2 C-5 Appendices
 
-Unit 1 . . . . . . . . . . . . . . . . . . . . .
Salem ODCM Rev. 6 03/28/90 Table of Contents - Continued 4.0 Radiological Environmental Monitoring Program .
41 Parameters for Liquid Alarm Setpoint Determination
* 36 4.1 Sampling Program . . . . . . . . . . . . . .
-Unit 2 . . . . . . . . .  
* 36 4.2 Interlaboratory Comparison Program         ... . *
* . . . 42 Site Related Ingestion Dose Commitment Factors, Aio . . . . . . . . .  
* 37 Tables 1-1. Parameters for Liquid Alarm Setpoint Determination
* . . . . . . .
              - Unit 1 .  . . . . . . . . .  .  . . . . . . . . . . 41 1-2 Parameters for Liquid Alarm Setpoint Determination
43 Bioaccumulation Factors (BFi) * . . . . 45 Dose Factors for Noble Gases . . . * . . . . . . .
              - Unit 2 . . . . . .         . . .* . . .             42 1-3  Site Related Ingestion Dose Commitment Factors, Aio . . . . . . . . . * . . . . . . .         43 1-4  Bioaccumulation Factors (BFi) * . . .               . 45 2-1  Dose Factors for Noble Gases . . . * . . . . . . . 48 2-2  Parameters for Gaseous Alarm Setpoint Determinations
48 Parameters for Gaseous Alarm Setpoint Determinations
              - Unit 1  .  . . . . . . . . . .  . . . . . . . . . . 49 2-3  Parameters for Gaseous Alarm Setpoint Determinati*ons
-Unit 1 . . . . . . . . . . . . . . . . . . . . . 4 9 Parameters for Gaseous Alarm Setpoint Determinati*ons
              - Unit 2 . . . . . . . . . . . . . . . . . . . . . 5 0 2-4 Controlling Locations,* Pathways and Atmospheric Dispersion for Dose Calculations        . . . . . . . 51
-Unit 2 . . . . . . . . . . . . . . . . . . . . .
.-        2-5 A-1 A-2 B-1 B-2 Pathway Dose Parameters - Atmospheric Releases . . 52 Calcu~ation of Effective MPC - Unit 1 * . . . . . . A-4 Calculation of Effective MPC - Unit 2 . . . . . . . A-5 Adult Dose Contributions Fish and Drinking Water Pathways Unit 1 * * . . * . . . . . . . . . . .
5 0 Controlling Locations,*
Adult Dose Contributions Fish and Drinking Water
Pathways and Atmospheric Dispersion for Dose Calculations . . . . . . .
                                                                    . B-5 Pathways Unit 2    * . * * . * . . * . . . . . . . . B-5 C-5  Effective Dose Factors . * *        . . . . . . . . . C-5 Appendices Appendix A - Evaluation of Conservative, Default MPC Value for Liquid Effluents * * . * . . . A-1 Appendix B - Technical Basis for Effective Dose Factors -
51 Pathway Dose Parameters
Liquid Radioactive Effl~ents * . . . . . B-1 Appendtx C - Technical Bases for Effective Dose Factors -
-Atmospheric Releases . . 52 of Effective MPC -Unit 1 * . . . . . .
Gaseous Radioactive Effluents        . . . C-1 Appendix D - Radiological Environmental Monitoring Program -
A-4 Calculation of Effective MPC -Unit 2 . . . . . . .
Sample Type, Location and Analysis . . . D-1
A-5 Adult Dose Contributions Fish and Drinking Water Pathways Unit 1 * * . . * . . . . . . . . . . . .
 
B-5 Adult Dose Contributions Fish and Drinking Water Pathways Unit 2 * . * * . * . . * . . . . . . . .
Salem ODCM Rev. 6 03/28/90 SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL 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
B-5 Effective Dose Factors . * * . . . . . . . . .
                                                      - -
C-5 Appendix A -Evaluation of Conservative, Default MPC Value for Liquid Effluents
this manual is acceptable for use in demonstrating compliance with 10 CFR 20.106, 10 CFR 50, Appendix I and 40 CFR 190.
* * . * . . . A-1 Appendix B -Technical Basis for Effective Dose Factors -Liquid Radioactive
More conservative calculation methods and/or conditions (e.g.,
* . . . . .
location and/or exposure pathways) expected to yield higher
B-1 Appendtx C -Technical Bases for Effective Dose Factors -Gaseous Radioactive Effluents . . . C-1 Appendix D -Radiological Environmental Monitoring Program -Sample Type, Location and Analysis . . . D-1 Introduction Salem ODCM Rev. 6 03/28/90 SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL 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 50, Appendix I and 40 CFR 190. More conservative calculation methods and/or conditions (e.g.,
* computed doses than appropriate for the maximally exposed person may be assumed in the dose evaluations.
* 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 methodologies and parameters as is deemed necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR 50.36a and Appendix I for demonstrating radioactive effluents are ALARA.
The ODCM will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations.
NOTE:  As used throughout this document, excluding acronyms, words appearing all capitalized denote the application of definitions as used in the Salem Technical Specifications.
Changes will be made to the ODCM calculation methodologies and parameters as is deemed necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR 50.36a and Appendix I for demonstrating radioactive 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
1
*
 
* Salem ODCM Rev. 6 03/28/90 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 (and Automatic Termination)  R18 (Unit 1) and 2-R18 (Unit 2) provide the alarm and automatic termination of liquid radioactive material releases as required by Technical Specification 3.3.3.8. 1-R19 A,B,C,and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R19 A,B,C and D provide this function for Unit 2 . 2) Alarm Conly) -The alarm functions for the Service Water 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.
Salem ODCM Rev. 6 03/28/90
* 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 (and Automatic Termination) R18 (Unit 1) and 2-R18 (Unit 2) provide the alarm and automatic termination of liquid radioactive material releases as required by Technical Specification 3.3.3.8.
1-R19 A,B,C,and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines.
2-R19 A,B,C and D provide this function for Unit 2 .
* 2)  Alarm Conly) - The alarm functions for the Service Water 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 1 and 2, respectively.
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 1 and 2, respectively.
2
~-
-. Salem ODCM Rev. 6 03/28/90 1.2 Liquid Effluent Monitor setpoint Determination Per the requirements of Technical Specification 3.3.3.8, alarm setpoints shall be established for the liquid 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 Ir,* Column 2, for radionuclides and 2.0E-04 uCi/ml for dissolved or entrained noble gases). The following equation*
2
must be satisfied to meet the liquid effluent restrictions:
 
where: C (F+f) (1.1) c f C = the effluent concentration limit of Technical 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
-. 1.2 Salem ODCM Rev. 6 03/28/90 Liquid Effluent Monitor setpoint Determination Per the requirements of Technical Specification 3.3.3.8, alarm setpoints shall be established for the liquid eff~uent 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 Ir,* Column 2, for radionuclides and 2.0E-04 uCi/ml for dissolved or entrained noble gases). The following equation* must be satisfied to meet the liquid effluent restrictions:
* 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, c< c. Also, note that when (F) is large compared to (f), then (F + f) = F.] Adapted from NUREG-0133 3
C (F+f)                   (1.1) c ~
Salem ODCM Rev. 6 03/28/90 1.2.1 Liquid Effluent Monitors (Radwaste, 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:
f where:
C = the effluent concentration limit of Technical 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, c< c. Also, note that when (F) is large compared to (f), then (F + f) = F.]
Adapted from NUREG-0133 3
 
Salem ODCM Rev. 6 03/28/90 1.2.1   Liquid Effluent Monitors (Radwaste, 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
MPCe
* SEN
* SEN
* CW SP + bkg ( 1. 2) RR with: MPCe = ------------( 1. 3) where: C* l. --------MPCi SP = alarm setpoint corresponding to the maximum allowable release rate (cpm) _ MPCe = an effective MPC value for the mixture of radionuclides in the effluent stram (uCi/ml) Ci = the concentration of radionuclide i in the undiluted liquid effluents (uCi/ml)*
* CW SP ~                        + bkg                       ( 1. 2)
MPCi = SEN = cw = RR = bkg = *NOTE The concentration mix must include the most recent composite of alpha emitters, sr-89, Sr-90, Fe-55 1 and H-3 per Technical Specification 3.11.1.1.
RR with:
the MPC value corresponding 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 (gal/min) the liquid effluent release rate (gal/min) the background of the monitor (cpm) 4
MPCe   = ------------                                   ( 1. 3)
* Salem ODCM Rev. 6 03/28/90 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.
C*l.
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.
MPCi where:
lo2.2 Conservative Default Values. Conservative alarm setpoints may be determined through the use of default parameters.
SP     = alarm setpoint corresponding to the maximum allowable release rate (cpm)         _
Tables 1-1 and 1-2 summarize all current default values in use for Salem Unit-1 and Unit-2, respectively.
MPCe   = an effective MPC value for the mixture of radionuclides in the effluent stram (uCi/ml)
Ci   =   the concentration of radionuclide i in the undiluted liquid effluents (uCi/ml)*
                  *NOTE         The concentration mix must include the most recent composite of alpha emitters, sr-89, Sr-90, Fe-55 1 and H-3 per Technical Specification 3.11.1.1.
MPCi  =  the MPC value corresponding 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 (gal/min)
RR    =  the liquid effluent release rate (gal/min) bkg    =  the background of the monitor (cpm) 4
 
Salem ODCM Rev. 6 03/28/90 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.
lo2.2 Conservative Default Values. Conservative 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 the following:
They are based upon the following:
a) substitution of the effective MPC value with a default value of lE-05 uCi/ml for radwaste releases (refer to Appendix A for justification)  
a) substitution of the effective MPC value with a default value of lE-05 uCi/ml for radwaste releases (refer to Appendix A for justification) ;
; b) for additional conservatism*, substitution of the I-131 MPC value of 3E-07 uCi/ml for the R19 Steam Generator
b) for additional conservatism*, substitution of the I-131 MPC value of 3E-07 uCi/ml for the R19 Steam Generator
* Use of the effective MPC value as derived in Appendix A may be non-conservative for the R19 Steam Generator blowdown monitors and R37 Chemical Waste Basin monitors where I-131 transfer during primary to secondary leakage may potentially be more controlling.blowdown monitors, 1R13** Service Water monitor and R37 Chemical Waste Basin monitor; 5 Salem ODCM Rev. 6 03/28/90 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 MPCe and for differing Circulating Water System dilutions.  
* Use of the effective MPC value as derived in Appendix A may be non- conservative for the R19 Steam Generator blowdown monitors and R37 Chemical Waste Basin monitors where I-131 transfer during primary to secondary leakage may potentially be more controlling.blowdown monitors, 1R13** Service
** The Unit 2 Service Water system utilizes the Unit 1 Circulating Water system 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 2Rl3 default alarm setpoint.
* Water monitor and R37 Chemical Waste Basin monitor; 5
Because no dilution is considered and the 2Rl3 monitor sensitivity is high, the MPCe of lE-05 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. 6 03/28/90 1.3 Liquid Effluent concentration Limits -10 CFR 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 limits of Technical Specification 3.11.1.1 may be performed using the following equation:
Salem ODCM Rev. 6 03/28/90 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.
C* 1 ------where: C* 1 MPC* 1 MPCi RR cw = = = = = RR
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 MPCe and for differing Circulating Water System dilutions.
* 1 (1.4) actual concentration of radionuclide i as measured in the undiluted liquid effluent (uCi/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 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
** The Unit 2 Service Water system utilizes the Unit 1 Circulating Water system 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 2Rl3 default alarm setpoint. Because no dilution is considered and the 2Rl3 monitor sensitivity is high, the MPCe of lE-05 uci/ml is used in calculating the alarm setpoint (otherwise using 3E-07 uCi/ml would result in an alarm setpoint of 1 cpm).
* Salem ODCM Rev. 6 03/28/90 1.4 Liquid Effluent Dose Calculation  
6
-10 CFR so 1.4.1 MEMBER OF THE PUBLIC Dose -Liquid Effluents.
 
Technical Specification 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; 1.5 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. 1. 67E-02
Salem ODCM Rev. 6 03/28/90 1.3   Liquid Effluent concentration Limits - 10 CFR 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 limits of Technical Specification 3.11.1.1 may be performed using the following equation:
* VOL cw where:
C*1               RR
* I: (C *
    ~  ------         *               ~  1                      (1.4)
* A* ) 1 10 (1.5) 0 0 = dose or dose commitment to organ o, including total body (mrem) Aio = site-related ingestion dose commitment factor to the total body or any organ o for radionuclide i (mrem/hr per uCi/ml) Ci = average concentration of radionuclide i, in undiluted
MPC*1         ~+RR where:
* 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. 6 03/28/90 The site-related ingestion dose/dose commitment factors (Aio> are presented in Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:  
C*1       =  actual concentration of radionuclide i as measured in the undiluted liquid effluent (uCi/ml)
= 1.14E+05 [(UI
MPCi      =-- 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. 6 03/28/90
* 1.4 1.4.1 Liquid Effluent Dose Calculation - 10 CFR so MEMBER OF THE PUBLIC Dose - Liquid Effluents.
Technical Specification 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;
            ~  1.5 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.
: 1. 67E-02
* VOL
* I: (C 1*
* A*10 )           (1.5) cw where:
00    = dose or dose commitment to organ o, including total body (mrem)
Aio = site-related ingestion dose commitment factor to the total body or any organ o for radionuclide i (mrem/hr per uCi/ml)
Ci = 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. 6 03/28/90 The site-related ingestion dose/dose commitment factors (Aio> are presented in Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:
        = 1.14E+05 [(UI
* Bii) + (UF
* Bii) + (UF
* BFi)J DFi ( 1. 6) where: Aio = 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) DFi = dose conversion factor for nuclide i for adults in pre-selected organ, o, from Table E-11 of Regulatory Guide 1.109 (mrem/pCi) l.14E+05 = conversion factor (pCi/uCi
* BFi)J DFi           ( 1. 6) where:
* ml/kg per hr/yr) The radionuclides included in the periodic dose assessment per the requirements of Technical Specification 3/4.11.1.2 are those 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. Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90) will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of Technical Specification Table 4.11-1. 9
Aio = 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)
* Salem ODCM Rev. 6 03/28/90 1.4.2 simplified Liquid Effluent Dose Calculation.
Bii = bioaccumulation factor for radionuclide i in invertebrates from Table 1-4 (pCi/kg per pCi/l)
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
UF = adult fish consumption (21 kg/yr)
* VOL = * ( 1. 7) cw Maximum Organ where: C* l. 2.52E+04
BFi = bioaccumulation factor for radionuclide i in fish from Table 1-4 (pCi/kg per pCi/l)
* VOL = * ( 1. 8) cw = 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)
DFi = dose conversion factor for nuclide i for adults in pre-selected organ, o, from Table E-11 of Regulatory Guide 1.109 (mrem/pCi) l.14E+05 = conversion factor (pCi/uCi
Dtb = conservatively evaluated total body dose (mrem) Dmax = 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, LLI --1.51E+06 mrem/hr per uci/ml) 10
* ml/kg per hr/yr)
** * ** Salem ODCM Rev. 6 03/28/90 1.s Secondary Side Radioactive Liquid Effluents and Dose Calculations During 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 discharges.
The radionuclides included in the periodic dose assessment per the requirements of Technical Specification 3/4.11.1.2 are those 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.
Potentially significant radioactive material levels and potential releases are controlled/monitored by the Steam Generator blowdown monitors (R19) 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.
Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90) will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of Technical Specification Table 4.11-1.
With identified radioactive material levels in the secondary system, appropriate samples should be collected and analyzed for the principal gamma __ emitting radionuclides.
9
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. 6 03/28/90 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.
Salem ODCM Rev. 6 03/28/90
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.
* 1.4.2     simplified Liquid Effluent Dose Calculation.
Accounting for radioactive material retention of the condensate clean-up system ion exchange resins may be needed to more accurately account for actual releases.
the individual radionuclide dose assessment as presented in In lieu of 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.)
12
Total Body
: 1. 21E+03
* VOL
                =                                               ( 1. 7) cw
* Maximum Organ
                                *
* 2.52E+04   VOL
                =                     *                       ( 1. 8) cw where:
C*l.    = 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)
Dtb = conservatively evaluated total body dose (mrem)
Dmax = 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 Rev. 6 03/28/90 1.s Secondary Side Radioactive Liquid Effluents and Dose Calculations During 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 discharges.
Potentially significant radioactive material levels and potential releases are controlled/monitored by the Steam Generator blowdown monitors (R19) 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. 6 03/28/90
* 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. 6 03/28/90 1.6 Liquid Effluent Dose Projections 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.
Salem ODCM Rev. 6 03/28/90 1.6 Liquid Effluent Dose Projections 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:
These processing requirements are applicable to each unit individually.
0.375 mrem to the total body, or 1.25 mrem to any organ.
Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing requirements for the other unit. Dose projections are made at least once per 31 days by the following equations:
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.
Dtbp 0 maxp = = Dtb (91 I d) Dmax (91 I d) 13 ( 1. 9) ( 1. 10) where: Dtbp Dtb 0 maxp 0 max d 91 ** = = =
These processing requirements are applicable to each unit individually. Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing requirements for the other unit.
= = = Salem ODCM Rev. 6 03/28/90 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 
Dose projections are made at least once per 31 days by the following equations:
** ----
Dtbp   =   Dtb (91 I d)                       ( 1. 9) 0 maxp  =  Dmax (91 I d)                     ( 1. 10) 13
Salem ODCM Rev. 6 03/28/90 2.0 Gaseous Effluents 2.1 Radiation 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 Rev. 6 03/28/90 where:
The tanks are discharged via the Plant Vent. 1-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 1-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-R12A and 2-R12A may be used to provide the containment monitoring and automatic isolation function during purge and pressure/vacuum reliefs.*
Dtbp  = the total body dose projection for current calendar quarter (mrem)
: 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.
Dtb    = the total body dose to date for current calendar quarter as determined by Equation 1.5 or 1.7 (mrem) 0 maxp = the maximum organ dose projection for current calendar quarter (mrem) 0 max  = 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
Effluents are monitored by R41C, a flow through gross activity monitor (for noble gas monitoring).
**                                   14
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.
Salem ODCM Rev. 6 03/28/90 2.0   Gaseous Effluents 2.1   Radiation 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:
The R41C monitor may also provide this function if the R12A monitor is inoperable during MdDE-6 . 15
: 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.
* Salem ODCM Rev. 6 03/28/90 3) Plant Vent (cont'd) R16) provide redundant back-up monitoring capabilities to the R41C monitors.
Waste gas decay tanks are batch released after sampling and analysis. The tanks are discharged via the Plant Vent.
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.
1-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.
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 rotometer).
: 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 1-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-R12A and 2-R12A may be used to provide the containment monitoring and automatic isolation function during purge and pressure/vacuum reliefs.*
: 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 MdDE-6 .
15
 
Salem ODCM Rev. 6 03/28/90
: 3) Plant Vent (cont'd) R16) 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 rotometer).
A 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.
A 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 Salem ODCM Rev. 6 03/28/90 2.2 Gaseous Effluent Monitor Setpoint Determination 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 release rate are used to calculate the fraction of the allowable release rate, as limited by Specification 3.11.2.1, by the equation:
16
FRAC = [4.72E+02
* Salem ODCM Rev. 6 03/28/90 2.2   Gaseous Effluent Monitor Setpoint Determination 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 release rate are used to calculate the fraction of the allowable release rate, as limited by Specification 3.11.2.1, by the equation:
* X/Q *VF* L (Ci* Ki)]/ 500 ( 2. 1) FRAC = [4.72E+02
FRAC = [4.72E+02
* X/Q *VF* L (Ci* (Li+ 1.1 Mi))] / 3000 (2.2) where: --FRAC = fraction of the allowable release rate based on the identified radionuclide concenrations and the release flow rate X/Q VF C* l. K* l. = = = = 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 determined by radioanalysis of grab sample (uCi/cm 3) total body dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) 17
* X/Q *VF*     L (Ci* Ki)]/ 500             ( 2. 1)
.. ----------Salem ODCM Rev. 6 03/28/90 L* 1 = beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m3 from Table 2-1) M* 1 = gamma air dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad) 500 = total body dose rate limit (mrem/yr) 3000 = skin dose rate limit (mrem/vr) 4.72E+02 = conversion factor (cm 3/ft 3
FRAC = [4.72E+02
* min/sec) Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoints for the applicable monitors (Rl6; R41C, and/or Rl2A) may be calculated by the equation:
* X/Q *VF* L (Ci* (Li+ 1.1 Mi))] /       3000 (2.2) where: --
SP = [AF
FRAC = fraction of the allowable release rate based on the identified radionuclide concenrations and the release flow rate X/Q   = annual average meteorological dispersion to the controlling site boundary location (sec/m 3 )
* Ci
VF    =  ventilation system flow rate for the applicable release point and monitor (ft 3 /min)
* SEN / FRAC] + bkg (2.3) where: SP = SEN = bkg = AF = 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 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 imposed to ensure that combined releases from Salem Units 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).
C*l.  =  concentration of noble gas radionuclide i as determined by radioanalysis of grab sample (uCi/cm 3 )
Normally, the combined AF value for Salem Units 1 and 2 is 0.5 (0.25 per unit), with the remainder o.s 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 18
K*l.  =  total body dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) 17
* Salem ODCM Rev. 6 03/28/90 allocation factors for all units do not exceed 1.0. 2.2.2 Conservative 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*
      -----
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 on 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.
Salem ODCM Rev. 6 03/28/90
* Adopted from ANSI N237-1976/ANS-18.1, Source Term Specifications, Table 6 19
.
'
L*1   = beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m3 from Table 2-1)
* Salem ODCM Rev. 6 03/28/90 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations  
M*1   = gamma air dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad) 500 = total body dose rate limit (mrem/yr) 3000 = skin dose rate limit (mrem/vr) 4.72E+02 = conversion factor (cm3 /ft 3
-10 CFR 20 2.3.1 Site 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 5500 mrem/yr, total body and 53000 mrem/yr, skin. Radiation monitor alarm setpoints are established to ensure that these release limits are not exceeded.
* min/sec)
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 Ds = X/Q * (2.5) where: Dtb Ds X/Q Qi K* 1 L* 1 M* 1 1.1 = =
Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoints for the applicable monitors (Rl6; R41C, and/or Rl2A) may be calculated by the equation:
= = = = = = 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) 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) gamma 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)
SP = [AF * ~ Ci
As appropriate, simultaneous releases from Salem Units 1 and 2 and Hope Creek will be considered in evaluating compliance with the release rate limits of Specification 3.11.2.la, following any 20 Salem ODCM Rev. 6 03/28/90 release exceeding the above prescribed alarm setpoints.
* SEN / FRAC] + bkg                   (2.3) where:
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.
SP   = alarm setpoint corresponding to the maximum allowable release rate (cpm)
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.
SEN = monitor sensitivity (cpm per uCi/cm 3 )
As identified, any electronic spiking monitor responses may be excluded from the analysis.
bkg = background of the monitor (cpm)
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.
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 imposed to ensure that combined releases from Salem Units 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 0.5 (0.25 per unit), with the remainder o.s 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 18
 
Salem ODCM Rev. 6 03/28/90
* allocation factors for all units do not exceed 1.0.
2.2.2 Conservative 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* 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 on 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 Specifications, Table 6 19
 
Salem ODCM Rev. 6 03/28/90 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations - 10 CFR 20 2.3.1 Site 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 5500 mrem/yr, total body and 53000 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 Ds = X/Q *                                       (2.5)
'  where:
Dtb     =  total body dose rate (mrem/yr)
Ds       =  skin dose rate (mrem/yr)
X/Q     =  atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m 3 )
Qi      =   average release rate of radionuclide i over the release period under evaluation (uCi/sec)
K*1      =  total body dose conversion factor for noble gas radionuclide i (mrem/yr per uci/m 3 , from Table 2-1)
L*1      =  beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 , from Table 2-1)
M*1     =  gamma air dose conversion factor for noble gas
                    *radionuclide i (mrad/yr per uCi/m 3 , from Table 2-1) 1.1      =  mrem skin dose per mrad gamma air dose (mrem/mrad)
As appropriate, simultaneous releases from Salem Units 1 and 2 and Hope Creek will be considered in evaluating compliance with the release rate limits of Specification 3.11.2.la, following any 20
* Salem ODCM Rev. 6 03/28/90 release exceeding the above prescribed alarm setpoints. 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.
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.
Actual meteorological conditions concurrent with the release -period or the default, annual average dispersion parameters as presented in--Table 2-4 may be used for evaluating the gaseous effluent dose rate. 2.3.2 Site Boundary Dose Rate -Radioiodine and Particulates.
Actual meteorological conditions concurrent with the release
Technical Specification 3.11.2.1.b limits the dose rate to mrem/yr to any organ for I-131, tritium and particulates with 21 Salem ODCM Rev. 6 03/28/90 half-lives greater than 8 days. To demonstrate compliance with this 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:
                    -
where: Do = X/Q = Rio = Qi -. = X/Q * (2.6) average organ dose rate over the sampling time period (mrem/yr) atmospheric dispersion to the controlling SITE BOUNDARY location for the inhalation pathway (sec/m31 dose parameter for radionuclide i (mrem/yr per uci/m ) I 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, tritium or other radionuclide in particulate form with half-life greater than 8 days (uCi/sec)
period or the default, annual average dispersion parameters as presented in- -Table 2-4 may be used for evaluating the gaseous effluent dose rate.
By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for I-131 can be determined.
2.3.2   Site Boundary Dose Rate - Radioiodine and Particulates.
Based on the annual average meteorological dispersion (see Table 2-4) and the most limiting potential pathway, age group and organ (inhalation, child, thyroid --Ri = 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 22
Technical Specification 3.11.2.1.b limits the dose rate to   ~1500 mrem/yr to any organ for I-131, tritium and particulates with 21
* Salem ODCM Rev. 6 03/28/90 release rate allocated to each of the Salem 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. 23
 
* * ** Salem ODCM Rev. 6 03/28/90 2.4 Noble Gas Effluent Dose Calculations  
Salem ODCM Rev. 6 03/28/90 half-lives greater than 8 days. To demonstrate compliance with this 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:
-10 CFR 50 2.4.1 UNRESTRICTED 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 dose limits of 55 mrad, gamma-air and 510 mrad, beta-air and the calendar year limits 510 mrad, gamma-air and 520 mrad, beta-air.
                = X/Q *                                   (2.6) where:
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
Do  = 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/m31 Rio = dose parameter for radionuclide i (mrem/yr per uci/m )
and organ o for the child inhalation pathway from I
Table 2-5 Qi -. average release rate over the appropriate sampling period and analysis frequency for radionuclide i --
I-131, I-133, tritium or other radionuclide in particulate form with half-life greater than 8 days (uCi/sec)
By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for I-131 can be determined. Based on the annual average meteorological dispersion (see Table 2-4) and the most limiting potential pathway, age group and organ (inhalation, child, thyroid -- Ri = 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 22
 
Salem ODCM Rev. 6 03/28/90
* release rate allocated to each of the Salem 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.
23
 
Salem ODCM Rev. 6 03/28/90
* 2.4   Noble Gas Effluent Dose Calculations - 10 CFR 50 2.4.1   UNRESTRICTED 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 dose limits of 55 mrad, gamma-air and 510 mrad, beta-air and the calendar year limits 510 mrad, gamma-air and 520 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
* L  (Mi
* Qi)                (2.7) and
              =  3.17E-08
* X/Q
* X/Q
* L (Mi
* L   (Ni
* Qi) and = 3.17E-08
* Qi)                 (2.8)
* where:
Dg
        ~
              =
              =
air dose due to gamma emissions for noble gas radionuclides (mrad) air dose due to beta emissions for noble gas radionuclides (mrad)
X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m 3 )
Qi = cumulative release of noble gas radionuclide i over the period of interest (uCi)
Mi = air dose factor due to gamma emissions from noble gas radionuclide i (mrad/yr per uci;m 3 , from Table 2-1)
Ni = air dose factor due to beta emissions from noble gas radionuclide i (mrad/yr per uCi/m 3 , Table 2-1) 3.17E-08 = conversion factor (yr/sec)
**                                      24
 
Salem ODCM Rev. 6 03/28/90 2.4.2    Simplified 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 3.11.2.2.        (Refer to Appendix c for the derivation and justification for this simplified method.)
3.17E-08 Dg    =  --------
a.so
* X/Q
* X/Q
* L (Ni
* Qi) where: Dg = air dose due to gamma emissions for noble gas radionuclides (mrad) = air dose due to beta emissions for noble gas radionuclides (mrad) X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m 3) (2.7) (2.8) Qi = cumulative release of noble gas radionuclide i over the period of interest (uCi) Mi = air dose factor due to gamma emissions from noble gas radionuclide i (mrad/yr per uci;m 3 , from Table 2-1) Ni = air dose factor due to beta emissions from noble gas radionuclide i (mrad/yr per uCi/m 3 , Table 2-1) 3.17E-08 = conversion factor (yr/sec) 24 Salem ODCM Rev. 6 03/28/90 2.4.2 Simplified 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 3.11.2.2. (Refer to Appendix c for the derivation and justification for this simplified method.) Dg = Db = where: Mef f = Neff = Qi = 0.50 = 3.17E-08 --------* X/Q
* Mef f
* Mef f
* L Qi (2.9) a.so and 3.17E-08 --------* X/Q
* L Qi               (2.9) and 3.17E-08 Db    =  --------
* X/Q
* Neff
* Neff
* L Qi (2.10) 0.50 5.3E+02, effective gamma-air dose factor (mrad/yr per uCi/m 3) 1.1E+03, effective beta-air dose factor (mrad/yr per uCi/m 3) cumulative release for all noble gas radionuclides (uCi) conservatism factor to account for potential variability in the radionuclide distribution . -Actual meteorological conditions concurrent with the release period or the default, annual average dispersion parameters as presented in may be used for the evaluation of the gamma-air and beta-air doses. 25 Salem ODCM Rev. 6 03/28/90 2.5 Radioiodine and Particulate Dose Calculations  
* L Qi               (2.10) 0.50 where:
-10 CFR 50 2.5.1 UNRESTRICTED AREA Dose -Radioiodine and Particulates.
Mef f =  5.3E+02, effective gamma-air dose factor (mrad/yr per uCi/m 3 )
In accordance with requirements of Technical Specification 3.11.2.3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit of 57.5 mrem and calendar year limit 515 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 a days: 0 aop where: = 0 aop w SFp = 3.17E-08
Neff  =  1.1E+03, effective beta-air dose factor (mrad/yr per uCi/m 3 )
Qi    = cumulative release for all noble gas radionuclides (uCi) 0.50  = conservatism factor to account for potential variability in the radionuclide distribution
                                                .-
Actual meteorological conditions concurrent with 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. 6 03/28/90 2.5   Radioiodine and Particulate Dose Calculations - 10 CFR 50 2.5.1   UNRESTRICTED AREA Dose - Radioiodine and Particulates.
In accordance with requirements of Technical Specification 3.11.2.3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit of 57.5 mrem and calendar year limit 515 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 a days:
0 aop   = 3.17E-08
* W
* W
* SFp * (Riop
* SFp *   ~  (Riop
* 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 identified in Table 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) dose factor for radionuclide i (mrem/yr per uCi/m 3) or (m 2 -mrem/yr per uCi/sec) and organ o from Table 2-5 for each age group a and the applicable pathway p as identified in Table 2-4. Values for Rio 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 greater than 8 days (uCi). annual seasonal correction factor to account for the fraction of the year that the applicable exposur_e 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:  
0 aop  = dose or dose commitment via all pathways p and controlling age group a (as identified in Table 2-
= 1.0 26
: 4) to organ o, including the total body (mrem) w      =  atmospheric dispersion parameter to the controlling location(s) as identified in Table 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)
* Salem ODCM Rev. 6 03/28/90 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.
            =  dose factor for radionuclide i (mrem/yr per uCi/m 3 )
2.s.2 Simplified Dose Calculation for Radioiodines and Particulates.
or (m2 - mrem/yr per uCi/sec) and organ o from Table 2-5 for each age group a and the applicable pathway p as identified in Table 2-4. Values for Rio were derived in accordance with the methods described in NUREG- 0133.             -   -
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). 0 max = where: 0 max = RI-131 = = w = Qi = 3.17E-08
            =  cumulative release over the period of interest for radionuclide i -- I-131 or radioactive material in particulate form with half- life greater than 8 days (uCi).
SFp  =  annual seasonal correction factor to account for the fraction of the year that the applicable exposur_e 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 Rev. 6 03/28/90
* 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. 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).
0 max   = 3.17E-08
* W
* W
* SFp
* SFp
* RI-131
* RI-131 * ~  Qi         ( 2 . 12)
* Qi ( 2 . 12) maximum organ dose (mrem) I-131 dose parameter for the thyroid for the identified controlling pathway l.05E+12, infant thyroid dose parameter with the cow-milk pathway controlling (m2 -mrem/yr per uCi/sec) D/Q for radioiodine, 2.lE-10 1/m 2 cumulative release over the period of interest for radionuclide i --I-131 or radiaoctive material in particulate from with half life greater than 8 days (uCi) The location of exposure pathways and the maximum organ dose 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  
* where:
.* pathways as identified in Table 2-4. 27 Salem ODCM Rev. 6 03/28/90 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.
0 max RI-131
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., R15 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., releases due to moisture carry over). Since this pathway is not sampled or monitored, it is necessary to calculate potential releases.
maximum organ dose (mrem)
Based on the guidance in NRC NUREG-0133, the releases of the radioiodines-
I-131 dose parameter for the thyroid for the identified controlling pathway
*and particulates shall be calculated by the equation:  
                  =  l.05E+12, infant thyroid dose parameter with the cow-milk pathway controlling (m2 - mrem/yr per uCi/sec) w      =  D/Q for radioiodine, 2.lE-10 1/m2 Qi      =  cumulative release over the period of interest for radionuclide i -- I-131 or radiaoctive material in particulate from with half life greater than 8 days (uCi)
= where: (2.13) = the release rate of radionuclide, i, from the steam flash tank vent (uCi/sec) 28 C* 1 Salem ODCM Rev. 6 03/28/90 = 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 SQftv = the measured steam quality in the flash tank vent; or an assumed value of 0.85, based on NUREG-0017.
The location of exposure pathways and the maximum organ dose 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
Tritium releases via the steam flashing may also be quantified using the above equation with the assumption of a steam quality (SQftv> 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 (pft) is approximately 0.48. The equation simplifies to the following:
.
(2.14) For H-3, the simplified equation is: (2.15) Also during reactor shutdown operations with a radioactively contaminated secondary
* pathways as identified in Table 2-4.
: system, material may be released to the atmosphere via the atmospheric reliefs (PORV) and 29
27
* * ** Salem ODCM_ Rev. 6 03/28/90 the safety 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: where: PF* 1 0.13 (2.16) = release rate of radionuclide i via pathway j (uCi/sec)  
Salem ODCM Rev. 6 03/28/90 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. 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., R15 on condenser evacuation, R41C on plant vent, and the plant vent particulate and charcoal samplers).
=-concentration of radionuclide i, in pathway j, (uCi/sec)  
However, if the Steam Generator blowdown is routed directly to the Chemical Waste Basin (via the SG blowdown flash tank) instead
= steam flow for release pathway j = 450,000 lb/hr per PORV = soo,ooo lb/hr per safety relief valve = __ 50, ooo lb/hr for auxiliary feed pump exhaust = partitioning factor, ratio of concentration in steam to that in the water in the steam generator  
  .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., releases due to moisture carry over). Since this pathway is not sampled or monitored, it is necessary to calculate potential releases.
= 0.01 for  
Based on the guidance in NRC NUREG-0133, the releases of the radioiodines- *and particulates shall be calculated by the equation:
= 0.005 for all other particulates  
          =                                                 (2.13) where:
= 1.0 for H-3 =conversion factor -[(hr*ml) / (sec*lb)]
          = the release rate of radionuclide, i, from the steam gener~tor flash tank vent (uCi/sec)
Any significant releases of noble gases via the atmospheric steam releases can be -quantified in accordance with the calculation methods of the Salem Emergency Plan Implementation Procedure . 30 I Salem ODCM Rev. 6 03/28/90 Alternately, the quantification of the release rate and cumulative releases may be based on actual samples of main steam collected at the R46 sample locations.
**                                    28
The measured radionuclide concentration in the steam may be used for quantifying the noble gases, radioiodine and particulate releases.
 
Salem ODCM Rev. 6 03/28/90 C*1    =   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 SQftv =   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 (SQftv> 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 (pft) is approximately 0.48. The equation simplifies to the following:
(2.14)
For H-3, the simplified equation is:
(2.15)
Also during reactor shutdown operations with a radioactively contaminated secondary system,   r~dioactive  material may be released to the atmosphere via the atmospheric reliefs (PORV) and 29
 
Salem ODCM_ Rev. 6 03/28/90
* the safety 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)       I where:
              = 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
            = soo,ooo lb/hr per safety relief valve
              = __ 50, ooo lb/hr for auxiliary feed pump exhaust PF*1 = partitioning factor, ratio of concentration in steam to that in the water in the steam generator
            = 0.01 for radioiodine~
            = 0.005 for all other particulates
            = 1.0 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 of the Salem Emergency Plan Implementation Procedure .
**                                      30
 
Salem ODCM Rev. 6 03/28/90 Alternately, the quantification of the release rate and cumulative releases may be based on actual samples of main steam collected at the R46 sample locations. The measured radionuclide concentration in the steam may be used for quantifying the noble gases, radioiodine and particulate releases.
Note: The expected mode of operation would be to isolate the effected steam generator, thereby reducing the potential releases during the shutdown/cooldown process. Use of the above calculation methods should consider actual operating conditions and release mechanisms.
Note: The expected mode of operation would be to isolate the effected steam generator, thereby reducing the potential releases during 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 demonstrating compliance with the Radiological Effluent Technical Specifications.
The calculated quantities of radioactive materials may be used as inputs to the equation (2.11) or (2.12) to calculate offsite doses for demonstrating compliance with the Radiological Effluent Technical Specifications.
31
31
* Salem ODCM Rev. 6 03/28/90 2.7 Gaseous Effluent Dose Projection 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:
Salem ODCM Rev. 6 03/28/90
where: Dgp Dg .Dmaxp Dmax d 91 Dgp = = Dmaxp = Dg * (91 I d) * (91 I d) Dmax * (91 I d) (2.17) (2.18) ( 2
* 2.7   Gaseous Effluent Dose Projection 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:
* 19) = = = = = = = = -gamma air dose projection for current calendar quarter (mrad) gamma air dose to date for current calendar quarter as by Equation 2.7 or 2.9 (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) 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
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.
* Salem ODCM Rev. 6 03/28/90 3.0 Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY 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 assessment of radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.
Dose projections are performed at least once per 31 days by the following equations:
There is one location on Artificial Island that is accessible to MEMBERS OF THE PUBLIC for activities unrelated to PSE&G operational and support activities.
Dgp   = Dg * (91 I d)                       (2.17)
This location is the Second Sun (visitor's center) located near the contractors gate for the Salem Nuclear Generating Station. 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 and 2 hours per visit per year. The default value for the meteorological dispersion data as in Table 2-3 may be used if current year meteorology is unavailable at the time of NRC reporting.
              ~p    =  ~  * (91 I d)                       (2.18)
Dmaxp  =  Dmax  * (91 I d)                     ( 2
* 19) where:                                          -
Dgp    = gamma air dose projection for current calendar quarter (mrad)
Dg      = gamma air dose to date for current calendar quarter as d.e~~rmined by Equation 2.7 or 2.9 (mrem)
        ~p      = 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)
      .Dmaxp  = maxi~um organ dose projection for current calendar quarter (mrem)
Dmax    = 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
 
.                                           Salem ODCM Rev. 6 03/28/90
* 3.0 3.1 Special Dose Analyses Doses Due To Activities Inside the SITE BOUNDARY 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 assessment of radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.
There is one location on Artificial Island that is accessible to MEMBERS OF THE PUBLIC for activities unrelated to PSE&G operational and support activities. This location is the Second
* Sun (visitor's center) located near the contractors gate for the Salem Nuclear Generating Station.
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 and 2 hours per visit per year. The default value for the meteorological dispersion data as pre~ented in Table 2-3 may be used if current year meteorology is unavailable at the time of NRC reporting.
However, a follow-up evaluation shall be performed when the data becomes available.
However, a follow-up evaluation shall be performed when the data becomes available.
33   
33
 
.                                          Salem ODCM Rev. 6 03/28/90 3.2    Total dose to MEMBERS OF THE PUBLIC - 40 CFR 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). 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 ~rom 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
* Salem ODCM Rev. 6 03/28/90 3.2.1 Effluent Dose 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.
3.2.2 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 would yield potentially significant off-site doses (i.e., in excess of 1 mrem per year to a MEMBER OF THE PUBLIC), that would require detailed evaluation for demonstrating compliance with 40 CFR 190.
However, should a situation exist whereby the direct exposure contribution is potentially significant, on-site measurements, off-site measure~ents and/or calculation techniques will be used determination of dose for assessing 40 CFR 190 compliance .
*
* for 35
 
Salem ODCM Rev. 6 03/28/90
* 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 iri 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 environment.
The sampling requirements (type of samples*, collection frequency and analysis) and sample locations are presented in Appendix E.
    *NOTE:  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
 
Salem ODCM Rev. 6 03/28/90 4.2 Interlaboratory comparison Program Technical Specification 3.12.3 requires analyses be performed on 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 Program results will be provided in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.10.
37
* RADIATION MONITOR                              QUID RELEASES UNIT 2 FIGURE 1-2 I        FROM VOl. CONIROI. IRNK 12 REACTOR LEIDOIN HX                                                              CCXUVH DARIN TIN:
I VENT TO GASEOUS J"
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                            -        JO HDLD-ll' IRllCS COll'OIENT COil.iNG SlflllE-TRllC REFUELING IATER STORAGE to-JANI(
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                        -
STEii! GENElllTDR IRSTE HONITOA HDLD-IP TfNC 12 J' IRSTE HDLD-11' TRllC 121    r r  lllSTE IO.IHI' TRllC m NO. 2 lllSTE EVll'ORATDR  f--o lllSTE HONITDR TRllC 122 121 EVll'IJllTDR FEED ION EXClfNlER  --            124 EVll'ORATOA FEED ION EXClllNGER 122 EVll'OfWITOR FEED ION EXCllNlER I+---
l 123 EVll'ORATOR FEED ION EXClflNGER SAlf'\.E Llll:S A19 HON ITCJISITJ STEii!
                      ..                                        STEii!                                                                        NO. 2 6flS GE NEiii TORS                                          BENElllTDR                                                                        STRIPPER                                                          SALEM OOCM 121.22.23.24            STEAlt GENElllTDR BllllllOIN  BLOIOOIN TRllC LINES                                                                                                                                                                        <FDR 11.-CJRMA TI ON ONLY l
[!]RIB NON-APO LIQUID IASTE DISPOSAL All r1 TO CIRCutATING IATER SYSTEH
__J DELRIARE AI VER
 
RADIATION MON                            G LIQUID RELEASES UNIT 1
                                                                                                                                                                                                                              '
I          fROl1 VOL. CIJN!AOl lnNK
                                                                                                                                                                                                                            ..
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                                                                  ;
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i VDLIK                                                                                                                  lllSTE 1111 ITlll CONTRll. TllNC    ~
TANC Ill Slff'LIHB SYSTE"        ~
lllSTE lllllTlll 111.D-lJ' TIN:
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                                                                                                                        ...__.  "'fTE lllllTlll IN: 112 E~Tlll FEED ION EXDINER 114 EVll'CRITlll FEED ION EXClllH6ER 112 EVll'lllATlll FEED lllf EXClllH6ER lo---
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-
R19 ITlllSITJ STEii! GENERATlll Sfff'LE LINES STElll GENEMTlllS 111.12.13,14        STElll GENEMTlll ILOIOOIN LINES STElll GENERllTlll ILOIOOIN TllNC HD* I &AS STRIPPER
(!] RIB NON-RAD LIOOID IASTE DISPOSll.
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_J                                                            SALEH OOCH If OR INflllltAI ION ONL YI DELRIFff RIVER
 
*N  vaporator package and/or r.adwaste demineralizer system LAUNDRY AND SllOWEa DllAINS LA* EQUIP. AND f'LOOA DaAINS      LA8 DRAINS          TO WASTE GAS COMPRESSOR IWL99      JWL91 LAUNDRY LAUNDRY        CHE*llCAL              GAS AND    ANO            DRAIN HOT                                          ANALYZ&#xa3;k HOT            TANK SHOW EA IHOWl.R aADWASTI:                                                                                                                            ACCUMULATOll UllJ\INS PACKAGE Ht:ttJl:lJNG EXCESS LETlJOW N CAN:\L
                                                                                                                                                                      ' ltlACTOlt t"LANGl. U:At..Ut t*
SOLID RADIOACTIVE                                                                          RWST                          LOUP UKAINS WASTll S\'STt:M eves (HIC)
                                                                                                          *.z
                                                                                                          ..<                                                    l'HESSUKllt.:H HJ::LU:r TANK 0
                                                                                                          .
Q
                                                                                                          ....
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                                                                                                              .,                  IWLl6          REACTOR COOLANT le    z                                DRAIN TANK PUMPS
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                                                                                                        ." ".
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                                                                                                        ... ..
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                                                                                                      *"
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Salem ODCM Rev. 6 03/28/90 Table 1-1 Parameters for Liquid Alarm setpoint Determinations Unit 1 Parameter      Actual      Default        Units                            Conments Value        Value MPCe          calculated    1E-05
* uCi/ml            calculated for each batch to be released MPCI-131      3E-07        N/A          uCi/ml            I-131 MPC conservatively used for SG blow-down and Service Water monitor setpoints Ci            measured      N/A          uCi/ml            taken from ganma spectral analysis of liquid effluent MPCi              as        N/A          uCi/ml            taken from 10 CFR 20,  Appendix B, Table II, determined                                  Col. 2.
SEN  1-R18        as      2.9E+07      cpm per uCi/ml    radwaste effluent  (Cs-137) determined 1-R19                  2.9E+07                        Steam Generator blowdown (Cs-137)
(A,B,C,D) 1-R13                  1.2E+08                        Service Water - Containment fan cooling (A,B,C,D,E)                                          (Cs-137) cw                as        1.85E+OS      gpm            Circulating Water System, single CW  ~
determined RR  1-R18        as          120        gpm            determined prior to release;    release rate determined                                  can be adjusted for Technical  Specification compliance 1-R19                    120                        Steam Generator blowdown rate per generator 1-R13                    2500                        Service Water flow rate for Containment fan coolers SP  1-R18
* calculated  4.4E+05(+bkg)  cpm            Default alarm setpoints; more conservative values may be used as deemed appropriate and 1-R19** calculated    1.3E+04C+bkg)                  desirable for ensuring regulatory compliance and for maintaining releases ALARA.
1-R13** calculated    2.6E+03(+bkg)
* Refer to Appendix A for derivation
** The MPC values of I-131 C3E-07 UCi/ml) has been used for derivation of the R19 Steam Generator blowdown and R13 Service Water monitor setpoints as discussed in Section 1.2.2.
41
 
Salem ODCM Rev. 6 03/28/90 Table 1-2 Parameters for Liquid Alarm Setpoint Determinations Unit 2 Parameter        Actual      Default        Units                              Conments Value          Value MP Ce          calculated    1E-05
* uCi/ml          calculate for each batch to be released MPCl-131        3E-07        N/A            uCi/ml          1-131 MPC conservatively used for SG blow-down, Service Water and Chemical Waste Basin monitor setpoints Ci              measured      N/A            uCi/ml          taken from ganma spectral analysis of liquid effluent MPCi                as      N/A            uCi/ml          taken from 10 CFR 20, Appendix B, Table II, determined                                    Col. 2.
SEN    2-R18        as      8.8E+07      cpm per uCi/ml    radwaste effluent    (Cs-137) determined 2-R19                  8.8E+07                        Steam Generator blowdown (Cs-137)
CA,B,C,D) 2-R13                  8.8E+07                        Service Water - Containment fan cooling CA,B,C)                                                (Cs-137)
R37                  8.8E+07                        Chemical Waste Basin (Cs-137) cw                as        1.85E+05        gpm            Circulating Water System, single CW P\.11'>
determined                                    (Note no CW P\.11'> in service for 2R13 monitor see section 1.2.2)
RR    2-R18        as            120          gpm            determined prior to release;      release rate determined                                    can be adjusted for Technical    Specification compliance 2-R19                      120                          Steam Generator blowdown rate per generator 2-R13                      2500                        Service water flow rate for Containment fan coolers R37                      300                          Chemical Waste Basin discharge SP  2-R18** calculated      8.0E+OSC+bkg)    cpm            Default alarm setpoints; more conservative values may be used as deemed appropriate and 2-R19*** calculated      3.9E+04C+bkg)                  desirable for ensuring regulatory compliance and for maintaining releases ALARA.
2-R13*** calculated      8.8E+02C+bkg)
R37****    calculated    1_.6E+04C+bkg)
* Refer to Appendix A for derivation
**    Actual calculated setpoint for 2-R18 (1.3E+06) is greater than the full scale monitor indicator, therefore, for conesrvatism the recommended-setpoint has been reduced to 3.0E+OS cpm
****      The MPC value of 1-131 (3E-07 uCi/ml) has been used for derivation of the R19 Steam generator blowdown and the R37 Chemical Waste Basin monitor setpofnts as discussed in Section 1.2.2
 
Salem ODCM Rev. 6 03/28/90 Table 1-3 Site Related Ingestion Dose Commitment Factors, Aio (mrem/hr per uci/ml)
Nuclide      Bone    Liver    T.Body    Thyroid  Kidney
      ------- ------- -------- ------- ------- ------- -------
Lung        _____
GI-LLI
_.._
H-3                2.82E-1  2.82E-1  2.82E-1  2.82E-1  2.82E-l  2.82E-l C-14      1. 45E+4 2.90E+3  2.90E+3  2.90E+3  2.90E+3  2.90E+3    2.90E+3 Na-24      4.57E-1  4.57E-1  4.57E-1  4.57E-1  4.57E-1  4.57E-l  4.57E-l P-32      4.69E+6  2.91E+5  1.81E+5                              5.27E+5 Cr-51                        5.58E+O  3.34E+O  1.23E+O  7.40E+O  1. 40E+3 Mn-54              7.06E+3  1. 35E+3          2.10E+3            2.16E+4 Mn-56              1. 78E+2  3.15E+l            2.26E+2            5.67E+3 Fe-55      5.11E+4  3.53E+4  8.23E+3                    1.97E+4  2.03E+4 Fe-59      8.06E+4  1. 90E+5  7.27E+4                    5.30E+4    6.32E+5 Co-57              1.42E+2  2.36E+2                              3.59E+3 Co-58              6.03E+2  1. 35E+3                              l.22E+4 Co-60              1. 73E+3  3.82E+3                                3.25E+4 Ni-63      4.96E+4  3.44E+3  1. 67E+3                              7.18E+2 Ni-65      2.02E+2  2.62E+l  1. 20E+l                              6.65E+2 cu-64              2.14E+2  1. 01E+2          5.40E+2            1. 83E+4 Zn-65      1. 61E+5 5.13E+5  2.32E+5            3.43E+5            3.23E+5 Zn-69      3.43E+2  6.56E+2  4.56E+l            4.26E+2            9.85E+l Br-82                        4.07E+O                              4.67E+O Br-83                        7.25E-2                                l.04E-l Br-84                        9.39E-2                              7.37E-7 Br-85                        3.86E-3 Rb-86              6.24E+2  2.91E+2                                l.23E+2 Rb-88              1.79E+O  9.49E-1                                2.47E-ll Rb-89              l.19E+O  8.34E-1                                6.89E-14 Sr-89      4.99E+3            1. 43E+2                              8.00E+2 Sr-90      1. 23E+5          3.01E+4                                3.55E+3 Sr-91      9.18E+l            3.71E+O                                4.37E+2 Sr-92      3.48E+l            1. 51E+O                              6.90E+2 Y-90      6.06E+O            1. 63E-l                              6.42E+4 Y-91m      5.73E-2            2.22E-3                                1.68E-l Y-91      8.88E+l            2.37E+O                                4.89E+4*
Y-92      5.32E-1            1. 56E-2                              9.32E+3 Y-93      l.69E+O            4.66E-2                                5.35E+4 Zr-95      1. 59E+l 5.llE+O  3.46_E+O          8.02E+O            l.62E+4 Zr-97      8.81E-1  1.78E-1  8.13E-2            2.68E-1            5.51E+4 Nb-95      4.47E+2  2.49E+2  1. 34E+2          2.46E+2            1.51E+6 Nb-97      3.75E+O  9.49E-1  3.46E-1            1.llE+O            3.50E+3
,, .Mo-99                1.2SE+2. 2.43E+l            2.89E+2            2.96E+2 Tc-99m    1.30E-2  3.66E-2  4.66E-1            5.56E-l  1.79E-2    2.17E+l Tc-101    1. 33E-2 1. 92E-2  1. 88E-1          3.46E-1  9.81E-3    5.77E-14 43
 
Salem ODCM Rev. 6 03/28/90
* 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              3.51E+O              l.15E+2          5.44E+3 Ru-106    1. 59E+3              2.01E+2              3.06E+3          1.03E+5 Rh-103m Rh-106 Ag-llOm    1.56E+3    1.45E+3    8.60E+2              2.85E+3          5.91E+5 Sb-124    2.77E+2    5.23E+O    1.10E+2    6.71E-1            2.15E+2  7.86E+3 Sb-125    1.77E+2    1. 98E+O  4.21E+l    1. 80E-1          1.36E+2  1. 95E+3 Te-125m    2.17E+2    7.86E+1    2.91E+l    6.52E+l  8.82E+2          8.66E+2 Te-127m    5.48E+2    1. 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    1.47E+2    3.20E+2  3.89E+3          4.69E+3 Te-129    2.54E+O    9.55E-1    6.19E-1    1. 95E+O  1. 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    1.17E+2    4.61E+l    9.91E+3  1.82E+2          1.01E+2 I-131      2.18E+2    3.12E+2    1.79E+2    1. 02E+5  5.35E+2          8.23E+l I-132      1.06E+l    2.85E+l    9.96E+O    9.96E+2  4.54E+l          5.35E+O I-133      7.45E+l    1. 30E+2  3.95E+l    1. 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. 3-2E+l  6.08E+l    2.24E+l    4.01E+3  9.75E+l            6.87E+l Cs-134    ~.84E+3    1. 63E+4  1. 33E+4              5.27E+3  1.75E+3  2.85E+2 Cs-136    7.16E+2    2.83E+3    2.04E+3              1.57E+3  2.16E+2  3.21E+2 Cs-137    8.77E+3    1. 20E+4  7.85E+3              4.07E+3  1.35E+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    7.85E+O    5.59E-3    2.30E-1              5.23E-3  3.17E-3  1. 39E+l Ba-140    1.64E+3    2.06E+O    1.08E+2              7.02E-1  1.18E+O  3.38E+3 Ba-141    3.81E+O    2.88E-3    1.29E-1              2.68E-3  1.63E-3  1.SOE-9 Ba-142    1.72E+O    1.77E-3    1. OSE-1              1. SOE-3  1.00E-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    3.43E+O    2.32E+O    2.63E-1              1.0SE+O            8.86E+3 Ce-143    6.04E-1    4.46E+2    4.94E-2              1.97E-1            1.67E+4 Ce-144    1.79E+2    7.47E+l    9.59E+O              4.43E+l            6.04E+4 Pr-143    5.79E+O    2.32E+O    2.87E-1              1. 34E+O          2.54E+4 Pr-144    1. 90E-2  7.87E-3    9.64E-4              4.44E-3            2.73E-9 Nd-147    3.96E+O    4.58E+O    2.74E-1              2.68E+O            2.20E+4 W-187      9.16E+O    7.66E+O    2.68E+O                                  2.51E+3 Np-239    3.53E-2    3.47E-3    1. 91E-3              1.08E-2            7.11E+2 44
 
Salem ODCM Rev. 6 03/28/90
* Table 1-4 Bioaccumulation Factors (BFi)
(pCi/kg per pCi/liter)*
Element            Saltwater Fish                Saltwater Invertebrate H                    9.0E-01                          9.3E-Ol c                    l.8E+03                          1. 4E+03 Na                  6.7E-02                          1.9E-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                  l.OE+02                          l.OE+03 Ni                  1.0E+02                          2.5E+02 Cu                  6.7E+02                          l.7E+03 Zn                  2.0E+03                          5.0E+04 Br                  1.5E-02                          3.lE+OO Rb                  8.3E+OO                          l.7E+Ol Sr                  2.0E+OO                          2.0E+Ol y                    2.5E+Ol                          l.OE+03 Zr                  2.0E+02                          8.0E+Ol Nb                  3.0E+04                          l.OE+02 Mo                  1. OE+Ol                        l.OE+Ol Tc                  1. OE+Ol                        5.0E+Ol Ru                  3.0E+OO                          1. OE+03 Rh                    1. OE+Ol                        2.0E+03 Ag                    3.3E+03                          3.3E+03 Sb                  4.0E+Ol                          5.4E+OO Te                  1. OE+Ol                        l.OE+02 I                    l.OE+Ol                          5.0E+Ol Cs                    4.0E+Ol                          2.5E+Ol Ba                    1. OE+Ol                        l.OE+Ol La                    2.5E+Ol                          1. OE+03 Ce                    l.OE+Ol                          6.0E+02 Pr                    2.5E+Ol                          l.OE+03 Nd                    2.5E+Ol                          l.OE+03 w                    3.0E+Ol                          3.0E+Ol Np                    1. OE+Ol                        l.OE+Ol
* Values in this 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
 
*                                                                                                                                                    ~
                                                                                                                                                      '=RADIATION PW AREA RADIATION MONITOR
                                                                                                                                                                                **
                                                                                                                        ......
MIA...._
                                                                                                                        ............
                                                                                                                        . .IC      AAlClE
                                                                                                                                                  -+-GASEOUS EFFWENT
                                                                                                                                                  -----* El.EC1RICAL
                                                                                                                                                  --PNEUMATIC Al    ~  CDNWf.MHI
                * ,  ~  . . OOM: SI.Al. JAii.i.
                                                                                                                    ~    .....
lt!f ....
lllQ.l~R~:'Dt                                                                                          I
                                                                                                                        ..........
MIC
              --8"~~4~
1t11F.3 CXlH~NI                                                                                                    *...
::M      CONIMMNt MIWG 1 r;--;i L~*              ..,.
I
___ _J
                                                                                                                                                      ----------.
I
                                                                                                                                          .____ - - - - - ____
                                                                                                                +~1--f!Pl--~H CONDEfrtSIR AIR Al...,,..._    SYSJIM
_,
r                                                          FOR INFORMATION ONLY AWl Ill.DO AOOf r.;;.~~---                                    lllG41 I  '~'RC*CiASOECA't'~-"'-----'
I :!\~~1 ~                                                                            FIGURE 2-1
                                                                                    ~r.;;;:;;-----
s--    ~~                            M~:J~-I
_,...
PlAHl\'fHlS
  --...........                                                                                                                                                            RADIATION L-- - - - - - --'
    ...... .......    .......
                        .......
                        ....... .......  -.......
                                                                ~-
L-------1
                                                                          =*i.o--,  I
                                                                                              - - - --,
                                                                                        "&deg;:ou""""' ...~I L - - - - --::221 MONITORING GASEOUS
                        .......
M*Jlt
                        ........
M*lll
                                                  ~l
__l.IL_
Pll&#xa3; &OCO
                                                            ..Bm..ll!L a::Ml\JTIER
__JJL_                                                                                              1UNIT
    -.........        ... .,..
Mto1Sl AAIOIM
                                                  * ..c Ml
                                                  ....
MIC.--
IMIO
                                                  ....
                                                                  -*
                                                              ""101"5 RAJ1r1
::::--:::::....
                                                              .......
RA~
AA~
Mr        M4111
 
*                                                                                                                                                ~
                                                                                                                                                  'PROCESS RADIATION MONITOff Rt.I AREA RADIATION
                                                                                                                                                                                        **
                                                                                                                      ..... _                          MONITOff
                                                                                                                                              -+GASEOUS Effl.lJENl
                                                                                                  ...... ~                                    ------ B.ECmlCAl.
            .
                                                                                              .-.1*H1.DG            -            H JUI      . . SC M    I O --
                                                                                                                                              ----~nc
                  ~....-                  ...                                                            I  "'""
            . . ~ NJDIW ..... , , . . ,
I ..... ......ilClANE
            -~ ..- - , .........                  _0-                                                              ~ ... a.-
            -~ "~ Mlftltll **a'                                                                                    ~MIC      -10"9
                                                                                                                      *-...,
                                                                                                                              '-...
r--::
                                                                                                                                      ,        I~*
                                                                                                                                            ~-            Full .,._,.,
                                                                                                                                                  ""--'"""""" _J r:e:----;r-
                                                                    ~-*      ~*
L-------.....1        --- _J I
                                                                                                                        ...
                                                                                                                                      . __ _ _._    _l ~-
                                                                                                                                                  ...
                                                                                                                                                      ----------i
____
    -..,. .....................,,,....... ...--... M-............__
Ma
              ... om
  .JIM&.. .IEB.Jll.
                ""Git
                                          *uc
                                                    . . 1000 JB!ll&. ...IDllllG.
                                                    .._.,
                                                                    ~..- -o;;;;;.;.,
                                                                                  ""'?'
L - _...::::::_
l
                                                                                                                                      ~---          - --- -
                                                                                                                                              ------------,I
__.
FOR INFORMATION ONLY
    -...
M          M*m
_,                  =--    __......                                                                      ..... I
                                                                                                                                                - S f ( GAS~            STSYl'M
      -
      .....
                  ...
MQll MCJID
                ..._                      - ...-
                                          ..IC 1&deg;'50IMDI
                                                                                                                                            ....
                                                                                                                                      ~--livl_._ ~
fR{ . . GAS Ma DIECAT ~SI
                                                                                                                                                                  ....... MID A
I FIGURE 2-2
      --
MU          Md:ID
      ...        MCJll                                                                                                                                                            I
                ........_
      ...        MUJI RADIATION
_
at        M.Qtt
                                                                                        -,                                                                                            MONITORING IMIA
      ....
      ...
MC
                  --
MIGlll MIOIM M tDIM GASEOUS 2UNIT
 
Salem ODCM Rev. 6 03/28/90
*
* Table 2-1 Dose Factors for Noble Gases Total Body                                  Galllll8 Air        Beta Air Dose Factor        Skin Dose Factor        Dose Factor          Dose Factor Radionuclide        Ki                    Li                        Mi                Ni Cmrem/yr per uCi/m3) Cmrem/yr per uCi/m3)  Cmrad/yr per uCi/m3) Cmrad/yr per uCi/m3)
  ------------ -------------------- --------------------  -------------------  --------------------
Kr-83m            7.56E-02                                    1.93E+01            2.88E+02 Kr*85m            , .17E+03            1.46E+03              1.23E+03            1.97E+03 Kr-85            1.61E+01            1.34E+03              1.nE+01              1.95E+03 Kr-87            5.92E+03            9.73E+03              6. 17E+03            1.03E+04 Kr-88            1 .47E+04            2.37E+03              1.52E+04            2.93E+03 Kr-89            1.66E+04            1.01E+04              1. 73E+04            1.06E+04 Kr-90            1.56E+04            7.29E+03              1.63E+04            7.83E+03 Xe* 131m          9. 15E+01            4.76E+02              1.56E+02            1.11E+03 Xe-133m          2.51E+02            9.94E+02              3.27E+02            1.48E+03 Xe-133            2.94E+02            3.06E+02              3.53E+02            1.05E+03 Xe-135m          3. 12E+03            7. 11E+02              3.36E+03            7.39E+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. 6 03/28/90
* Table 2-2 Parameters for Gaseous Alarm Setpoint Determinations Unit-1 Parameter        Actual          Default          Units                      Cooments Value              Value
    --------      --------          --------      ----------      -------------------------------------------
X/Q            calculated        2.2E-06        sec/m3          USNRC Salem Safety Evaluation, Sup. 3 VF            as measured        1.25E+05      ft 3/min        Plant Vent - normal operation (Plant              or Vent)      fan curves VF              as measured      3.5E+04                        Containment purge (Cont.            or Purge)      fan curves AF              coordinated          0.25        unit less        Acministrative allocation factor to.*
with HCGS                                        ensure conbined re.Leases do not exceed release rate Limit for site.
Ci              measured          N/A            uCi/cm3 Ki          nuclide specific*- N/A          mrem/yr per uCi/m3  Values from Table 2-t Li          nuclide specific    N/A        mrem/yr per uCi/m3  Values from Table 2-1 Mi          nuclide specific    N/A        mrad/yr per uCi/m3  Values from Table 2-1 SEN 1-R41C*      as              1.6E+07    cpm per uCi/cm3    Plant Vent determined 1-R16                      3.6E+07                        Plant Vent (redundant)
                                      -
1-R12A                      2.1E+06                        Containment SP 1-R41C      calculated        3.3E+04C+blcg) cpm              Default alarm setpoints; more conservative values may be used as deemed appropriate and 1-R16    calculated        7.4E+04C+blcg)                  desirable for ensuring regulatory coq:>liance and for maintaining releases ALARA.
1-R12A** calculated        1.5E+04(+blcg)
* Based on mean.for calibration with mixture of radionuclides
  ** Applicable during MODES 1 through 5. During MODE 6 (refueling), monitor setpoint shall be reduced to 2X baclcgrourid in accordance with Tech Spec Table 3.3-6.
49
 
Salem ODCM Rev. 6 03/28/90 Table 2*3 Parameters for Gaseous Alarm Setpoint Determinations Unit-2 Parameter        Actual        Default          Units                        Comments Value            Value
    --------      --------                                        -------------------------------------------
X/Q            calculated      2.2E-06        sectm3            Licensing technical specification value*
VF            as measured      1.25E+OS                        Plant Vent - normal operation CPL ant            or Vent)      fan curves VF            as measured      3.SE+04                        Containment purge (Cont.              or Purge)      fan curves AF              coordinated        0.25        unitless          Administrative allocation factor to with HCGS                                      ensure coat>ined releases do not exceed release rate Limit for site.
Ci          measured            N/A          uCi/cm3
* Ki Li Mi nuclide specific nuclide specific nuclide specific N/A N/A N/A mrem/yr per UCitm3 mrem/yr per uCitm3 mrad/yr per uCitm3 Values from Table 2-1 Values from Table 2-1 Values from Table 2-1 SEN 2-R41C*      as            1.6E+07    cpm per uCi/cm3    Plant Vent determined 2-R16                      3.SE+07                        Plant Vent (redundant) 2-R12A                    3.3E+07                        Containment SP 2-R41C      calculated        3.3E+04C+bkg) cpm              Default alarm setpoints; more conservative values may be used as deemed appropriate and 2-R16    calculated        7.2E+04(+bkg)                  desirable for ensuring regulatory c~Liance and for maintaining releases ALARA.
2-R12A** calculated        2.4E+OSC+bkg)
* Based on mean for calibration with mixture of radionuclides
  ** Applicable during MOOES 1 through 5. During MOOE 6 (refueling), monitor setpoints shall be reduced to 2X background in accordance with Tech Spec Table 3.3-6.
50
 
Salem ODCM Rev. 6 03/28/90 Table 2-4 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations
* Atmospheric Dispersion Technical Specification          Location              Pathway(s)      Controlling      X/Q          D/Q Age Group      (sec/m3 >      C1tm2 >
                                                                                --------
3.11.2.1a            site boundary        noble gases            N/A        2.2E-06        N/A (0.83 mi le, N)      direct exposure 3.11.2.1b            site boundary          inhalation          child        2.2E-06        N/A (0.83 mile, N) 3.11.2.2            site boundary        gamma-air              N/A        2.2E-06        N/A (0.83 mi le, N)      beta-air 3.11.2.3            residence/dairy C4.9 miles, W) milk, ground plane and inhalation infant      S.4E-08      2.1E-10 I
6.9.1.10            Second sun            direct exposure        N/A        8.22E-06        N/A
*
(0.21 mile/SE)        and inhalation
* The identified controlling locations, pathways and atmospheric dispersion are from the Safety Evaluation Report, Supplement No. 3 for the Salem Nuclear Generating Station; Unit 2 CNUREG-0517, December 1978).
51
 
-
Salem ODCM Rev. 6 03/28/90
* 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    1.18E+4      2.78E+4      -          -    1. 02E+6 l.88E+5  l.06E+4 Co-57        -        6.92E+2      -          -    3.70E+5  3.14E+4  6.71E+2 Co-58        -        1.58E+3      -          -    9.28E+5  1. 06E+5  2.07E+3 Co-60        -        1.15E+4      -          -    5.97E+6  2.85E+5  1. 48E+4 Ni-63      4.32E+5      3.14E+4      -          -    1.78E+5  l.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      -          -        -    1.66E+4  5.90E+4 Sr-89    3.04E+5          -        -          -    1. 40E+6 3.50E+5  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    1. 07E+5      3.44E+4      -      5.42E+4  1.77E+6  1. 50E+5  2.33E+4 Nb-95      1. 41E+4      7.82E+3      -      7.74E+3  5.05E+5  l.04E+5  4.21E+3 Ru-103    1.53E+3          -        -      5.83E+3  5.05E+5  1.10E+5  6.58E+2 Ru-106    6.91E+4          -        -      1. 34E+5  9.36E+6  9.12E+5  8.72E+3 Ag-llOm    l.08E+4      1.00E+4      -      1. 97E+4  4.63E+6  3.02E+5  5.94E+3 Sb-124  - 3.12E+4      5.89E+2  7.55E+l        -    2.48E+6  4.06E+5  1. 24E+4 Sb-125 Te-125m 5.34E+4 3.42E+3 5.95E+2 1.58E+3 5.40E+l 1.05E+3
                                                      -    1.74E+6  1. 01E+5  1.26E+4 4.67E+2 1.24E+4  3.14E+5  7.06E+4 Te-127m    1.26E+4      5.77E+3  3.29E+3    4.58E+4  9.60E+5  1.50E+5  1.57E+3 Te-129m    9.76E+3      4.67E+3  3.44E+3    3.66E+4  1.16E+6  3.83E+5  l.58E+3 I-131    2.52E+4      3.58E+4  1.19E+7    6.13E+4      -    6.28E+3  2.05E+4 Cs-134    3.73E+5      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  l.10E+5 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      1.35E+4      -      6.26E+3  3.62E+5  1.20E+5  1.53E+3 Ce-144    3.43E+6      1.43E+6      -      8.48E+5  7.78E+6  8.16E+5  1.84E+5 Pr-143    9.36E+3      3.75E+3      -      2.16E+3  2.81E+5  2.00E+S  4.64E+2
".-Nd-147      5.27E+3      6.lOE+3      -      3.56E+3  2.21E+5  1. 73E+5  3.65E+2 52
 
..
..
* Salem ODCM Rev. 6 03/28/90 3.2 Total dose to MEMBERS OF THE PUBLIC -40 CFR 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).
Salem ODCM Rev. 6 03/28/90 Table 2-5 (cont'd)
For the likely most exposed MEMBER OF THE PUBLIC in the vicinity of Artificial Island, the sources of exposure need only consider the
R ( io) , Inhalation Pathway Dose Factors - TEENAGER (mrem/yr per uCi/m3)
Nuclide  Bone        Liver    Thyroid  Kidney    Lung    GI-LLI  T.Body
    ------- ------- ------- ------- ------- ------- -
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
** where: Salem ODCM Rev. 6 3/28/90 C* 1 E --------MPCi = an effective MPC value for a mixture of radionuclide (uCi/ml), (A. l) = concentration of radionuclide i in the mixture = the 10 CFR 20, Appendix B, Table II, Column 2 MPC value for radionuclide i (uCi/ml) Based on the above equation and the radionuclide distribution in the effluents for past years from Salem, an effective MPC value can be determine.
 
Results are presented in Table A-1 and A-2 for Unit 1 and Unit 2, respectively.
Salem ODCM Rev. 6   3/28/90 (A. l)
Considering the average effective MPC value for the years 1981 through 1989, it is reasonable to select an MPCe value of lE-05 uCi/ml as typical of liquid radwaste discharges.
C*1 E --------
Using this value to calculate the default Rl8 alarm setpoint value, results in a setpoint that: 1) 2) Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations, and 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-3 4 *
MPCi where:
* Nuclide MPC CuCi/ml) --------Na-24 3E-05 Cr-51 2E-03 Mn-54 1E-04 Fe-59 5E-05 Co-58 9E-05 Co-60 3E-05 Zr-95 6E-05 Nb-95 1E-04 Nb-97 9E-04 Tc-99m 3E-03 Sr-89
              = an effective MPC value for a mixture of radionuclide (uCi/ml),
* 3E-06 Sr-90 3E-07 Mo-99 4E-05 Ag-110m 3E-05 Sn-113 8E-05 Sb-124 2E-05 Sb-125 1E-04 1-131 3E-07 1-133 1E-06 1-135 4E-06 Cs-134 9E-06 Cs-137 2E-05 Ba-140 2E-05 La-140 2E-05 Total Ci c. -1 MPCi MPCe (UCi/ml)
              = concentration of radionuclide i in the mixture
* Table A-1 Calculation of Effective MPC Salem Unit 1 Salem ODCM Rev. 6 3/28/90 Activity Released (Ci) ---------------------------------------------------------------------------------
              = the 10 CFR 20, Appendix B, Table II, Column 2 MPC value for radionuclide i (uCi/ml)
1984 1985 1986 1987 1988 1989 ----------------------------------------------------5.6E-03 6.2E-03 9.2E-04 6.9E-04 1.38E-02 4.69E-04 5.3E-02 3.6E-02 6.0E-02 N/D 2.38E-02 5.25E-03 1.9E-01 8.7E-02 1.9E-01 1.0E-01 1.01E-01 1.12E-01 5.8E-03 1.4E-03 2.4E-03 N/D 2.66E-04 1.32E-03 1.6 6.6E-01 2.22 1.54 1.27E+OO 1.82E+OO 1.2 6.5E-01 3.1E-01 4.2E-01 2.nE-01 1.78E-01 1.8E-03 3.2E-03 4.3E-03 8.6E-04 1.23E-02 1.53E-03 1. 7E-02 1.3E-03 1.8E-02 2.4E-03 1.53E-02 3.85E-03 2.0E-02 7.2E-03 1.5E-03 9.8E-03 2.44E-02 7.94E-05 1.6E-03 N/D N/D 1.1E-04 4.74E-03 4.62E-04 4.2E-04 1. 7E-03 3.5E-07 1.6E-02 1.25E-02 9.37E-04 2.2E-05 1. 7E-04 3.1E-08 7.7E-04 2.40E-03 3.75E-04 1.9E-03 1.0E-04 N/D 1.0E-04 1.57E-03 N/D N/D N/D N/D 2.8E-03 4.96E-03 2.70E-03 9.4E-04 N/D 3;5E-04 N/D N/D N/D 1.7E-02 5.7E-03 8.4E-02 2.4E-02 6.32E-02 1.36E-02 4.9E-03 N/D 3.6E-02 3.3E-02 9.35E-02 6.53E-02 4.5E-02 7.9E-02 1.2E-01 1.8E-01 5.54E-02 3.04E-02 1.9E-02 1.4E-03 1.9E-02 2.80E-02 6.88E-03 1.2E-03 N/D N/D 2.0E-03 1.68E-02 1.94E-04 5 .1E-02 1.6E-01 3.4E-01 3.1E-03 1.31E-01 1.16E-01 5.8E-02 2.1E-01 3.6E-01 3.0E-01 1.34E-01 1.28E-01 2.1E-03 N/D N/D N/D 2.79E-04 N/D 1.6E-02 1.1E-04 3.5E-04 N/D 3.89E-04 2.66E-04 3.32 1.93 3.75 3.26 2.29E+OO 2.49E+OO 2.46E+05 3.42E+05 4.99E+05 7.31E+05 2.80E+05 1.58E+05 1.35E-05 5.63E-06 7.51E-06 4.46E-06 8.18E-06 1.58E-05 MPC value for unrestricted area from 10 CFR 20, Appendix B, Table 11, Coluin 2. ** N/D -not detected A-4
Based on the above equation and the radionuclide distribution in the effluents for past years from Salem, an effective MPC value can be determine. Results are presented in Table A-1 and A-2 for Unit 1 and Unit 2, respectively.
*
Considering the average effective MPC value for the years 1981 through 1989, it is reasonable to select an MPCe value of lE-05 uCi/ml as typical of liquid radwaste discharges. Using this value to calculate the default Rl8 alarm setpoint value, results in a setpoint that:
* Nuclide MPC (uCi/ml) --------Na-24 3E-05 Cr-51 2E-03 Mn-54 1E-04 Fe-59 5E-05 Co-58 9E-05 Co-60 3E-05 Zr-95 6E-05 Nb-95 1E-04 Nb-97 9E-04 Tc-99m 3E-03 Sr-89 3E-06 Sr-90 3E-07 Mo-99 4E-05 Ag-110m 3E-05 Sn-113 8E-05 Sb-124 2E-05 Sb-125 1E-04 I-131 3E-07 1-133 1E-06 I-135 4E-06 Cs-134 9E-06 Cs-137 2E-05 Ba-140 2E-05 La-140 2E-05 Total Ci c. -1 MPCi MPCe (uCi/ml) Table A-2 Calculation of Effective MPC Salem Unit 2 Salem ODCM Rev. 6 3/28/90 Activity Released CCi) -----------------------------------------------------------------------------------
: 1) Will not require frequent re-adjustment due to minor variations in the nuclide distribution which are typical of routine plant operations, and
1984 1985 1986 1987 1988 1989 -----------*----------------------------------------4.4E-03 3.5E-03 3.6E-03 7.3E-05 1.04E-02 8.08E-04 3.6E-02 3.5E-02 9.5E-02 3.0E-03 3.17E-03 1.57E-02 1.6E-01 1.1E-01 2.2E-01 1.2E-01 1. 74E-01 1.19E-01 7.6E-03 1.1E-03 4.0E-03 N/D 2.93E-05 3.00E-03 1.3 8.4E-01 3.32 1.7 1.32E+OO 2.02E+OO 9.8E-01 6.3E-01 3.8E-01 4.2E-01 2.97E-01 2.08E-01 1.2E-03 4.6E-03 1.1E-02 8.4E-04. 3.15E-03 3.39E-03 1.4E-02 1.4E-02 2.5E-02 6.6E-03 6.55E-03 7.41E-03 2.1E-02 5.7E-03 2.7E-03 N/D 6.92E-03 2.54E-04 1.4E-03 N/D N/D 5.7E-04 3.28E-03 6.64E-04 3.2E-04 1.5E-03 --4.1E-07 3.0E-03 1.69E-02 9.22E-04 4.1E-05 1.0E-04 3.2E-08 2.9E-04 4.11E-03 2.71E-04 1.4E-03 N/D N/D 4.4E-04 1.19E-04 N/D N/D N/D N/D N/D 1.04E-02 6.41E-03 1.2E-03 N/D 1.1E-03 N/D N/D N/D 3.0E-02 1.2E-03 1.2E-01 4.6E-02 5.47E-02 1.89E-02 3.6E-03 N/D 5.4E-02 5.9E-02 9.22E-02 8.08E-02 4.2E-02 8.4E-02 1.2E-01 2.2E-01 1.35E-01 3.79E-02 2.6E-02 1.2E-02 2.6E-03 1.8E-02 8.83E-02 8.64E-03 4.4E-04 N/D N/D N/D 1.90E-02 5.17E-04 2.6E-02 1.8E-01 3.6E-01 3.5E-01 9.53E-02 1.43E-01 4.8E-02 2.3E-01 3.7E-01 3.3E-01 1.09E-01 1.55E-01 6.6E-03 N/D N/D N/D 1.57E-03 N/D 3.0E-02 N/D 6.9E-04 N/D 1.03E-03 5 .19E"04 2.74 2.15 5.09 3.85 2.45E+OO 2.83E+OO 2.24E+05 3.56E+05 5.20E+05 8.59E+05 6.09E+05 1.93E+05 1.22E-05 6.04E-06 9.79E-06 4.49E-06 4.02E-06 1.47E-05
: 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).
* MPC value for unrestricted area from 10 CFR 20, Appendix B, Table II, Colurn 2. ** N/D -not detected A-5
**                              A-3
 
Salem ODCM Rev. 6          3/28/90
* 4 Table A-1 Calculation of Effective MPC Salem Unit 1 Activity Released (Ci)
                          ---------------------------------------------------------------------------------
Nuclide    MPC
* 1984          1985            1986        1987        1988        1989 CuCi/ml)
              --------     ---------     ---------       ---------     ---------   --------     --------
Na-24      3E-05        5.6E-03      6.2E-03        9.2E-04      6.9E-04    1.38E-02    4.69E-04 Cr-51      2E-03        5.3E-02      3.6E-02        6.0E-02          N/D    2.38E-02    5.25E-03 Mn-54      1E-04        1.9E-01      8.7E-02        1.9E-01      1.0E-01    1.01E-01    1.12E-01 Fe-59      5E-05        5.8E-03      1.4E-03        2.4E-03          N/D    2.66E-04    1.32E-03 Co-58      9E-05        1.6          6.6E-01        2.22          1.54      1.27E+OO    1.82E+OO Co-60      3E-05        1.2          6.5E-01        3.1E-01      4.2E-01    2.nE-01     1.78E-01 Zr-95      6E-05        1.8E-03       3.2E-03         4.3E-03       8.6E-04   1.23E-02    1.53E-03 Nb-95      1E-04        1. 7E-02      1.3E-03        1.8E-02      2.4E-03    1.53E-02    3.85E-03 Nb-97      9E-04        2.0E-02      7.2E-03        1.5E-03       9.8E-03   2.44E-02    7.94E-05 Tc-99m    3E-03         1.6E-03          N/D            N/D        1.1E-04    4.74E-03     4.62E-04 Sr-89
* 3E-06        4.2E-04      1. 7E-03       3.5E-07      1.6E-02   1.25E-02    9.37E-04 Sr-90      3E-07        2.2E-05      1. 7E-04        3.1E-08      7.7E-04   2.40E-03     3.75E-04 Mo-99      4E-05        1.9E-03       1.0E-04           N/D       1.0E-04   1.57E-03       N/D Ag-110m    3E-05            N/D           N/D             N/D       2.8E-03   4.96E-03     2.70E-03 Sn-113    8E-05        9.4E-04         N/D         3;5E-04         N/D         N/D         N/D Sb-124    2E-05        1.7E-02       5.7E-03         8.4E-02       2.4E-02   6.32E-02     1.36E-02 Sb-125    1E-04        4.9E-03         N/D         3.6E-02       3.3E-02   9.35E-02     6.53E-02 1-131      3E-07        4.5E-02       7.9E-02         1.2E-01       1.8E-01   5.54E-02     3.04E-02 1-133      1E-06        1.9E-02       1.4E-03                       1.9E-02   2.80E-02     6.88E-03 1-135      4E-06        1.2E-03           N/D           N/D       2.0E-03   1.68E-02     1.94E-04 Cs-134    9E-06        5 .1E-02     1.6E-01         3.4E-01       3.1E-03   1.31E-01     1.16E-01 Cs-137    2E-05        5.8E-02       2.1E-01         3.6E-01       3.0E-01   1.34E-01     1.28E-01 Ba-140    2E-05        2.1E-03           N/D           N/D           N/D     2.79E-04       N/D La-140    2E-05        1.6E-02       1.1E-04         3.5E-04         N/D     3.89E-04     2.66E-04 Total Ci                  3.32           1.93           3.75         3.26     2.29E+OO     2.49E+OO c.
        -1 2.46E+05     3.42E+05       4.99E+05     7.31E+05   2.80E+05     1.58E+05 MPCi MPCe (UCi/ml)          1.35E-05     5.63E-06       7.51E-06     4.46E-06   8.18E-06     1.58E-05
* MPC value for unrestricted area from 10 CFR 20, Appendix B, Table 11, Coluin 2.
    ** N/D - not detected A-4
~-
 
Salem ODCM Rev. 6            3/28/90 Table A-2 Calculation of Effective MPC Salem Unit 2 Activity Released CCi)
                          -----------------------------------------------------------------------------------
Nuclide        MPC
* 1984          1985            1986        1987        1988          1989 (uCi/ml)
                --------     ---------     --*------       ---------   --------- --------     --------
Na-24        3E-05        4.4E-03      3.5E-03          3.6E-03    7.3E-05  1.04E-02      8.08E-04 Cr-51        2E-03        3.6E-02      3.5E-02          9.5E-02    3.0E-03  3.17E-03      1.57E-02 Mn-54        1E-04        1.6E-01      1.1E-01          2.2E-01    1.2E-01  1. 74E-01    1.19E-01 Fe-59        5E-05        7.6E-03      1.1E-03          4.0E-03        N/D    2.93E-05      3.00E-03 Co-58        9E-05         1.3          8.4E-01          3.32        1.7      1.32E+OO      2.02E+OO Co-60        3E-05        9.8E-01      6.3E-01         3.8E-01     4.2E-01   2.97E-01     2.08E-01 Zr-95        6E-05        1.2E-03      4.6E-03         1.1E-02    8.4E-04. 3.15E-03     3.39E-03 Nb-95        1E-04        1.4E-02      1.4E-02          2.5E-02    6.6E-03  6.55E-03      7.41E-03 Nb-97        9E-04        2.1E-02      5.7E-03          2.7E-03       N/D    6.92E-03     2.54E-04 Tc-99m        3E-03         1.4E-03          N/D              N/D      5.7E-04  3.28E-03     6.64E-04 Sr-89        3E-06        3.2E-04      1.5E-03     -- 4.1E-07      3.0E-03   1.69E-02      9.22E-04 Sr-90        3E-07        4.1E-05      1.0E-04         3.2E-08    2.9E-04  4.11E-03     2.71E-04 Mo-99        4E-05         1.4E-03         N/D             N/D     4.4E-04   1.19E-04         N/D
* Ag-110m      3E-05            N/D           N/D             N/D         N/D   1.04E-02     6.41E-03 Sn-113        8E-05        1.2E-03         N/D           1.1E-03       N/D       N/D         N/D Sb-124        2E-05        3.0E-02       1.2E-03         1.2E-01   4.6E-02   5.47E-02     1.89E-02 Sb-125        1E-04        3.6E-03         N/D           5.4E-02   5.9E-02   9.22E-02     8.08E-02 I-131        3E-07        4.2E-02       8.4E-02         1.2E-01   2.2E-01   1.35E-01     3.79E-02 1-133        1E-06        2.6E-02       1.2E-02         2.6E-03   1.8E-02   8.83E-02     8.64E-03 I-135        4E-06        4.4E-04         N/D               N/D       N/D   1.90E-02     5.17E-04 Cs-134        9E-06        2.6E-02       1.8E-01         3.6E-01   3.5E-01   9.53E-02     1.43E-01 Cs-137        2E-05        4.8E-02       2.3E-01         3.7E-01   3.3E-01   1.09E-01     1.55E-01 Ba-140        2E-05        6.6E-03         N/D               N/D       N/D   1.57E-03         N/D La-140        2E-05        3.0E-02         N/D           6.9E-04       N/D   1.03E-03     5 .19E"04 Total Ci                      2.74           2.15             5.09       3.85   2.45E+OO     2.83E+OO c.
        -1 2.24E+05       3.56E+05         5.20E+05 8.59E+05   6.09E+05     1.93E+05 MPCi MPCe (uCi/ml)              1.22E-05       6.04E-06         9.79E-06 4.49E-06   4.02E-06       1.47E-05
* MPC value for unrestricted area from 10 CFR 20, Appendix B, Table II, Colurn 2.
  ** N/D - not detected A-5
~-


Salem ODCM Rev. 6 3/28/90 ...
Salem ODCM Rev. 6 3/28/90
...
* APPENDIX B Technical Basis for Effective Dose Factors Liquid Radioactive Effluent
* APPENDIX B Technical Basis for Effective Dose Factors Liquid Radioactive Effluent
* B-1 Salem ODCM Rev. 6 3/28/90 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.
* B-1
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 J.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.
Salem ODCM Rev. 6   3/28/90 APPENDIX B Technical Basis for Effective Dose Factors -
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 radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
Liquid Effluent Releases 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 J.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 the evaluation of the maximum organ dose, it is conservative to
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 radionuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
* ** Salem ODCM Rev. 6 3/28/90 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 dose factor of all the radionuclides evaluated.
For the evaluation of the maximum organ dose, it is conservative to
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 e911ations may be used: Total Body 1.67E-02
Salem ODCM Rev. 6   3/28/90 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 dose 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.
* VOL Dtb = where: Dtb A Fe-59,TB VOL C* l. cw 1. 67E-02
For evaluating compliance with the dose limits of Technical Specification 3.11.1.2, the following simplified e911ations may be used:
* 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)  
Total Body 1.67E-02
= conversion factor (hr/min) Substituting the value for the Fe-59 total body dose conversion factor, the equation simplifies to: 1.21E+03
* VOL Dtb =                     A Fe-59,TB *                   (B. 1)
* VOL Dtb = (B.2) cw -* B-3
* cw
* *
* where:
* Salem ODCM Rev. 6 3/28/90 Maximum organ l.67E-02
Dtb          = dose to the total body (mrem)
A Fe-59,TB    = 7.27E+04, total body ingestion dose conversion factor for Fe-59 (mrem/hr per uCi/ml)
VOL          = volume of liquid effluent released (gal)
C*l.          = 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:
1.21E+03
* VOL Dtb =         cw (B.2)
**                              -
* B-3
 
Salem ODCM Rev. 6 3/28/90
*
* Maximum organ l.67E-02
* VOL
* VOL
* A Nb-95,GI-LLI  
* A Nb-95,GI-LLI
---------------------------
                      ---------------------------
* E C* i (B.3) cw where: Dmax = maximum organ dose (mrem) A Nb-95,GI-LLI  
* E C*i   (B.3) cw where:
= 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
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 = --------------
* VOL Dmax = --------------
* (B.4) cw Tritium is not included in the limited analysis dose assessment for liquid releases, because the potential dose resulting from normal reactor releases is relatively negligible.
cw
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.4)
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.
Tritium is not included in the limited analysis dose assessment for liquid releases, because the potential dose resulting from normal reactor releases is relatively negligible.       The average annual tritium release from each Salem Unit is approximately 350 curies.
B-4 1989 Radio-RELEASE TBOOY nuclide (Ci) Dose Frac. MN-54 FE-55 FE-59 C0-58 C0-60 ZN-65 1.12E-01 3.98E-02 1.32E-03 1.82E+OO 1.78E-01 3.62E-04 NB-95 3.85E-03 AG-110M 2.70E-03 0.03 0.05 0.02 0.39 0.11 0.01 *
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.
* CS-134 1.16E-01 0.24 CS-137 1.28E-01 0.16 Total 2.40E+OO 1989 ISOTOPE RELEASE TBOOY MN-54 FE-55 FE-59 C0-58 C0-60 ZN-65 NB-95 (Ci) 1.19E-01 4.61E-02 3.00E-03 2.02E+OO 2.0BE-01 1.41E-04 7.41E-03 AG-110M 6.41E-03 Dose Frac. 0.02 0.05 0.03 0.37 0.11 * *
Furthermore, the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste system operation.
* CS-134 1.43E-01 0.25 CS-137 1.55E-01 0.16 Total 2.71E+OO
* B-4
* less than 0.01 /D = not detected GI-LLI LIVER Dose Dose Frac. 0.06 0.02 0.02 0.56 0.15
 
* 0.15 0.04 *
Table B-1 Adult Dose contributions Fish and Invertebrate Pathways Unit 1 1989                                       1988                                  1987 Radio-   RELEASE TBOOY   GI-LLI  LIVER        RELEASE TBOOY        GI-LLI  LIVER    RELEASE  TBODY    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   0.03    0.06  0.11        1.01E-01      0.01      0.03     0.03  1.05E-01   0.01      0.05    0.-04 FE-55    3.98E-02  0.05    0.02  0.19        5.44E-01      0.43      0.17      0.76  2.35E-01  0.16      0.11    0.43 FE-59    1.32E-03  0.02   0.02  0.03        2.66E-04      *        *
* Frac. 0.11 0.19 0.03 0.16 0.04 0.02 *
* N/D      *          *
* 0.25 0.21 GI-LLI LIVER Dose Frac. 0.05 0.02 0.04 0.47 0.13
* C0-58    1.82E+OO  0.39    0.56  0.16       1.27E+OO     0.17      0.24      0.03 1.54E+OO   0.17      0.42    0.05 C0-60    1.78E-01  0.11    0.15  0.04       2.m-01        0.10      0.14      0.02 4.21E-01  0.13      0.31    0.04 ZN-65    3.62E-04    0.01
* 0.22 0.07 *
* 0.02        5.49E-04      0.01
* Dose Frac. 0.09 0.18 0.06 0.14 0.04 . 0.01 *
* 0.01     N/D     *          *
* 0.26 0.21 Table B-1 Adult Dose contributions Fish and Invertebrate Pathways Unit 1 1988 RELEASE TBOOY GI-LLI LIVER (Ci) Dose Dose Dose 1.01E-01 5.44E-01 2.66E-04 1.27E+OO 2.m-01 5.49E-04 1.53E-02 4.96E-03 Frac. 0.01 0.43
* NB-95    3.85E-03
* 0.17 0.10 0.01 *
* 0.15
* 1.31E-01 0.17 1.34E-01 0.10 2.48E+OO Table B-2 Frac. 0.03 0.17
* 1.53E-02
* 0.24 0.14
* 0.36
* 0.36 0.05 *
* 2.44E-03
* Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1988 Frac. 0.03 0.76
* 0.08
* 0.03 0.02 0.01 *
* AG-110M  2.70E-03
* 0.08 0.06 RELEASE TBOOY (Ci) . Dose Frac. GI *LLI LIVER 1. 74E-01 4.69E-01 2.93E-05 1.32E+OO 2.97E-01 N/D 6.55E-03 1.04E-02 0.03 0.42
* 0.04
* 0.19 0.12 * *
* 4.96E-03
* 9.53E-02 0.14 1.09E-01 0.09 2.48E+OO Dose Frac. 0.07 0.16
* 0.05
* 0.29 0.18
* 2.36E-03
* 0.18 0.11 *
* 0.03
* Dose Frac. . 0.06 0.75
* CS-134    1.16E-01  0.24
* 0.04 0.02 * *
* 0.25       1.31E-01      0.17
* 0.07 0.06 1987 RELEASE TBODY Gl*LLI LIVER (Ci) Dose * *Dose Dose 1.05E-01 2.35E-01 N/D 1.54E+OO 4.21E-01 N/D 2.44E-03 2.36E-03 Frac. 0.01 0.16
* 0.08  3.11E-01  0.34
* 0.17 0.13 * *
* 0.26 CS-137    1.28E-01  0.16
* 3.11E-01 0.34 3.01E-01 0.19 2.92E+OO 1987 RELEASE TBODY (Ci) 1.20E-01 8.74E-01 N/D 1. 71E+OO 4.23E*01 N/D 7.92E-03 N/D 3.49E-01 3.33E-01 3.82E+OO Dose Frac. 0.01 0.39
* 0.21        1.34E-01      0.10
* 0.12 0.09 * *
* 0.06 3.01E-01  0.19
* 0.25 0.14 Frac. 0.05 0.11
* 0. 19 Total    2.40E+OO                              2.48E+OO                                2.92E+OO Table B-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1989                                      1988                                 1987 ISOTOPE  RELEASE   TBOOY   GI-LLI LIVER        RELEASE TBOOY        GI *LLI  LIVER  RELEASE  TBODY    GI-LL!  LIVER (Ci)   Dose     Dose     Dose         (Ci) . Dose        Dose    Dose      (Ci)    Dose      Dose    Dose Frac. Frac. Frac.                 Frac. Frac. Frac.               Frac. Frac. Frac.
* 0.42 0.31
MN-54    1.19E-01  0.02    0.05    0.09      1. 74E-01     0.03      0.07  . 0.06 1.20E-01  0.01      0.04    0.02 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    0.72 FE-59    3.00E-03   0.03   0.04    0.06        2.93E-05      *        *
* 0.08 0.03 *
* N/D      *         *
* Frac 0.-04 0.43
* C0-58    2.02E+OO   0.37    0.47    0.14        1.32E+OO    0.19      0.29    0.04 1. 71E+OO  0.12      0.31    0.02 C0-60    2.0BE-01   0.11    0.13    0.04        2.97E-01    0.12      0.18    0.02 4.23E*01   0.09    0.21    0.02 ZN-65    1.41E-04    *      *    . 0.01           N/D       *        *
* 0.05 0.04 * *
* N/D     *          *
* 0.26 0. 19 GI-LL! LIVER Dose Frac. 0.04 0.26
* NB-95    7.41E-03
* 0.31 0.21
* 0.22
* 0.18 * *
* 6.55E-03
* Dose Frac. 0.02 0.72
* 0.18
* 0.02 0.02 * -...-* .
* 7.92E-03
* o. 13 -0.09  
* 0.18    -...-* .
... Salem ODCM Rev. 6 3/28/90
AG-110M  6.41E-03
* APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent * ** C-1 Salem ODCM Rev. 6 3/28/90 ..
* 0.07
* APPENDIX C Technical Bases for Effective Dose Factors -Gaseous Radioactive Effluents overview 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.
* 1.04E-02
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 (Keff 1 Meff or Neff) times the total quantity of radioactive material released
* 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
* o. 13 CS-137    1.55E-01  0.16
* 0.21        1.09E-01    0.09
* 0.06 3.33E-01    0.14      *   - 0.09 Total    2.71E+OO                                2.48E+OO 3.82E+OO
* less than 0.01
/D = not detected
 
...                                 Salem ODCM Rev. 6 3/28/90
* APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent
  *
**                 C-1
 
Salem ODCM Rev. 6 3/28/90
..
* 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 (Keff 1 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.
* 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:
Determination of .Effective Dose Factors Effective dose transfer factors are calculated by the following equations:
where: Kef f K* 1 f
(C.1) where:
* 1 = = = (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
Kef f   = the effective total body dose factor due to gamma emissions from all noble gases released K*1      =  the total body dose factor due to gamma emissions from each noble gas radionuclide i released f 1*    =  the fractional abundance of noble gas radionuclide i relative to the total noble gas activity C-2
* * ** Salem ODCM Rev. 6 3/28/90 * (C.2) where: (L + 1.1 M)eff = the effective skin dose factor due to beta and gamma emissions from all noble gases released where: where: = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released (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 emissions from 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. 6   3/28/90
* ** Salem ODCM Rev. 6 3/28/90 2,800 Ci for 1985, 2,700 Ci for 1986, 1700 Ci for 1988, and 1500 Ci for 1989. Therefore, in order to provide 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 conservatism 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.2) where:
For evaluating compliance with the dose limits of Technical Specification 3.11.2.2, the following simplified equations may be used: 3.17E-08 Dg = --* --------* X/Q
(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 (C. 3) where:
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
* where:
(C.4)
                =   the effective air dose factor due to beta emissions
                    .from all noble gases released
                =   the air dose factor due to beta emissions from 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. 6   3/28/90 2,800 Ci for 1985, 2,700 Ci for 1986, 1700 Ci for 1988, and 1500 Ci for 1989. Therefore, in order to provide a reasonable basis for r
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 conservatism 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:
3.17E-08 Dg     =--* --------
o.so
* X/Q
* Mef f
* Mef f
* E Qi (C. 5) o.so and 3.17E-08 = --------* X/Q
* E Qi               (C. 5) and 3.17E-08
              =   --------
o.so
* X/Q
* Neff
* Neff
* E Qi (C. 6) o.so C-4 r 
* E Qi               (C. 6)
..
**                              C-4
* where: Dg = Db = X/Q = Mef f = Neff = Qi = 3.17E-08 = 0.50 = Salem ODCM Rev. 6 3/28/90 air dose due to gamma emissions for the cumulative release of all noble gases (mrad) air dose due to beta emissions for the cumulative release 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 constants, the dose calculational equations simplify to: Dg = 3.5E-05
 
* X/Q
Salem ODCM Rev. 6   3/28/90
* E Qi (C.7) and = 7.0E-05
.
* X/Q
* where:
* E Qi (C.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.  
Dg Db
-. C-5 Noble Gases -Total Body and Skin Radionuclide f.* 1 ------------
                  = air dose due to gamma emissions for the cumulative release of all noble gases (mrad)
Kr-85 0.01 Kr-88 0.01 Xe-133m 0.01 Xe-133 0.95 Xe-135 0.02 Total Noble Gases -Air Radionuclide f.* 1 ------------
                  = air dose due to beta     emissions for the cumulative release of all noble gases (mrad)
Kr-85 0.01 Kr-88 0.01 Xe-133m 0.01 Xe-133 0.95 Xe-135 0.02 Total Table C-1 Effective Dose Factors Total Body Effective Dose Factor Ke ff (mrem/yr per UCi/m 3) --------------------
X/Q      =  atmospheric dispersion to the controlling site boundary (sec/m3)
1.5E+02 2.5E+OO 3.0E+02 3.6E+01 4.8E+02 Gamma Air Effective Dose Factor "ef f (mrad/yr per uCitm 3) --------------------
Mef f    =  5.3E+02, .effective gamma-air dose factor (mrad/yr per uCi/m3)
1.5E+02 3.3E+OO 3.4E+02 3.8E+01 5.3E+02 Salem ODCM Rev. 6 Skin Effective Dose Factor (l+ 1.1 M>eff (mrem/yr per uCitm 3) 1.4E+01 1.9E+02 1.4E+01 6.6E+02 7.9E+01 9.6E+02 Beta Air Effective Dose Factor Neff (mrad/yr per uCitm 3) ---------------------
Neff      =  1.1E+03, effective beta-air dose factor (mrad/yr per uCi/m3)
2.0E+01 2.9E+01 1.5E+01 1.0E+03 4.9E+01 1.1E+03
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             - .
* Based on Noble gas distribution from ANSI N237-1976/ANSI-18.1, "Source Term Specifications." C-6 3/28/90
Combining the constants, the dose calculational equations simplify to:
* Salem ODCM Rev. 6 3/28/90 APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1 Salem ODCM Rev. 6 3/28/90 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.
              =   3.5E-05       X/Q                             (C.7)
This analysis was performed to provide a simplified method for 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 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.
Dg                *
Multiplication of the total release (i.e. cumulative activity for all radlonuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
* E Qi and
              =   7.0E-05       X/Q                             (C.8)
                          *
* E Qi 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-5
 
Salem ODCM Rev. 6 3/28/90 Table C-1 Effective Dose Factors Noble Gases - Total Body and Skin Total Body Effective              Skin Effective Radionuclide       f.*
1 Dose Factor                    Dose Factor Ke ff                    (l+ 1.1 M>eff (mrem/yr per UCi/m3 )        (mrem/yr per uCitm3 )
------------                     --------------------
Kr-85               0.01                                                1.4E+01 Kr-88               0.01                 1.5E+02                        1.9E+02 Xe-133m             0.01                2.5E+OO                        1.4E+01 Xe-133               0.95               3.0E+02                        6.6E+02 Xe-135               0.02               3.6E+01                        7.9E+01 Total                                    4.8E+02                       9.6E+02 Noble Gases - Air Gamma Air Effective         Beta Air Effective Radionuclide        f.*              Dose Factor                  Dose Factor 1
                                          "ef f                         Neff (mrad/yr per uCitm3 )        (mrad/yr per uCitm3 )
------------                      --------------------         ---------------------
Kr-85                0.01                                              2.0E+01 Kr-88                0.01                1.5E+02                        2.9E+01 Xe-133m              0.01                3.3E+OO                        1.5E+01 Xe-133              0.95                3.4E+02                        1.0E+03 Xe-135              0.02                3.8E+01                        4.9E+01 Total                                    5.3E+02                        1.1E+03
* Based on Noble gas distribution from ANSI N237-1976/ANSI-18.1,   "Source Term Specifications."
C-6
 
Salem ODCM Rev. 6 3/28/90
* APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1
 
Salem ODCM Rev. 6 3/28/90 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 ~etermining  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~resented 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. Multiplication of the total release (i.e. cumulative activity for all radlonuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.
D-2
D-2
* Salem ODCM Rev. 6 3/28/90 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.
Salem ODCM Rev. 6   3/28/90 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 X/Q D/Q Qi 3.17E-8 RI-131 Dmax = 3.17E-8-*
For evaluating compliance with the dose limits of Technical Specification 3.11.2.3, the following simplified equation may be used:
W
Dmax   = 3.17E-8-* W
* RI-131
* RI-131
* E Qi = = = = = = = = maximum organ dose (mrem) atmospheric dispersion parameters to the controlling location(s) as identified in Table 3.2-4. atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways (sec/m3) atmosperic deposition for vegetation, milk nad 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 1.05El2 (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 fo the overall negligible contribution of these pathways to D-3 Salem ODCM Rev. 6 3/28/90 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 maximize 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 exposure pathways and the maximum organ soe 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. D-4   
* E Qi where:
Dmax        =   maximum organ dose (mrem) w          =  atmospheric dispersion parameters to the controlling location(s) as identified in Table 3.2-4.
X/Q  =  atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways (sec/m3)
D/Q  =  atmosperic deposition for vegetation, milk nad 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
                    =  1.05El2 (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 fo the overall negligible contribution of these pathways to
* D-3
 
Salem ODCM Rev. 6 3/28/90 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 maximize 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 exposure pathways and the maximum organ soe 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 ODCM Rev. 6 3/28/90
..
..
* Target Organs Total Body Liver Thyroid Kidney Lung GI-LLI Salem ODCM Rev. 6 3/28/90 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 Fraction of Dose Contribution  
* 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 }2y Pathway Pathway Grass-Cow-Milk           0.92 Ground Plane             0.08 Inhalation
}2y Pathway Pathway Grass-Cow-Milk 0.92 Ground Plane 0.08 Inhalation
* D-5
* D-5 Ground Plane 0.15 0.14 0.15 0.15 0.02 0.15
 
* Salem ODCM Rev. 6 3/28/90 Appendix E Radiological Environmental Monitoring Program Sample Type, Location and Analysis E-1
Salem ODCM Rev. 6 3/28/90 Appendix E Radiological Environmental Monitoring Program Sample Type, Location and Analysis
.. .. Salem ODCM Rev. 6 3/28/90 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)
* E-1
ESS = Sediment RWA = Rain Water (Precipitation)
 
FPB = Beef SWA = Surface Water FPL = Green Leafy Vegetables VGT = Fodder Crops (Various)
.                                             Salem ODCM Rev. 6 3/28/90 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.
FPV = Vegetable (Various)
AIO = Air Iodine                 IDM = Immersion Dose (TLD)
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 off-site c = 2-3 miles off-site H = > 20 miles off-site 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 J
APT = Air Particulates           MLK = Milk ECH = Hard Shell Blue Crab       PWR = Potable Water (Raw)
* Salem ODCM Rev. 6 3/28/90 . 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. STATION CODE TABLE E-1 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 402 501 1001 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 ESS ESS ECH, ESF, ESS SWA IDM AIO, APT, IDM, WWA IDM Salem ODCM Rev. 6 3/28/90 *
ESF = Edible Fish               PWT = Potable Water (Treated)
* 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 f qarm FPV 7El 4.5 mi. SE of Horse Creek vent; 1 mi. W of Mad ESF, ESS, SWA 9El 5.0 mi. SW of vent IDM 11E2 5.0 mi. SW of vent IDM
ESS = Sediment                   RWA = Rain Water (Precipitation)
* 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 16El 4.1 mi. NNW of vent; Port Penn AIO, APT, IDM lFl 5.8 mi. N of vent; Fort Elfsborg AIO, APT, IDM _lF2 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. 6 3/28/90 TABLE E-1 (Cont'd) STATION CODE STATION LOCATION SAMPLE TYPES 2F6 7.3 mi. NNE of vent; Southern Training IDM Center 2F7 5.7 mi. NNE of vent; local farm MLK, VGT 3F2 5 .1 mi. NE of vent; Hancocks Bridge IDM Municipal Building 3F3 a. 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 7F2 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 IDM 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 MLK, VGT 14F2 6.6 mi. WNW of vent; Boyds Corner IDM 14F3 5.4 mi. WNW of vent; local farm FPV 15F3 5.4 mi. NW of vent IDM E-5 STATION CODE 16Fl 16F2 lGl lG3 2Gl 3Gl lOGl 16Gl 3Hl 3H3 3H5 Salem ODCM Rev. 6 3/28/90 TABLE E-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 17 mi. NE of vent; local farm 12 mi. SSW of vent; Smyrna, DE 15 mi. NNW of vent; Greater Wilmington Airport 32 mi. NE of vent; National Park, NJ 110 mi. NE of vent; Research and Testing 25 mi. NE of vent; local farm E-6 SAMPLE TYPES ESS, SWA IDM FPV IDM FPV IDM, MLK, VGT IDM IDM IDM AIO, APT, IDM FPL, FPV 
FPB = Beef                       SWA = Surface Water FPL = Green Leafy Vegetables     VGT = Fodder Crops (Various)
** Salem ODCM Rev. 6 3/28/90 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 A TEDA impregnated charcoal cartridge is connected to air particulated air sampler and is collected weekly at filter change. Crab and Fish Two batch samples are sealed in a plastic bag or jar and frozen semi-annually or when in season. Sediment A sediment sample is taken 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 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:
Direct 2 TLD's will be collected from each location quarterly.
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 off-site c = 2-3 miles off-site           H = > 20 miles off-site 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),
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 thoron daughter decay. Gamma isotopic analysis on quarterly composites.
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.
Iodine 131 analysis are performed on each weekly sample. Gamma isotopic analysis of edible portion on collection.
.                                E-2 J
Gamma isotopic analysis semi-annually.
 
Gamma dose quarterly 
Salem ODCM Rev. 6   3/28/90.
* 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      0.14 mi. SSW of vent; site shoreline   IDM llSl      0.09 mi. SW of vent; site shoreline     IDM llAl      0.2 mi. W of vent; outfall area         ECH, ESF, ESS, SWA 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 402      3.7 mi. ENE of vent; Alloway Creek     IDM Neck Road 501      3.5 mi. E of vent; local farm           AIO, APT, IDM, WWA 1001      3.9 mi. SSW of vent; Taylor's Bridge   IDM Spur E-3
 
Salem ODCM Rev. 6   3/28/90
*
*
* Sample Milk Water (Rain, Potable, Surface) Salem ODCM Rev. 6 3/28/90 SAMPLE COLLECTION AND ANALYSIS (Cont'd) Collection Method Sample of fresh milk is collected for each farm semi-monthly when cows are in pasture, monthly at other times. Sample to be collected monthly providing winter icing conditions allow. E-8 Analysis Gamma isotopic analysis and I-131 analysis on each sample on collection.
* STATION CODE TABLE E-1 (Cont'd)
Gamma isotopic*
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 f qarm            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
monthly H-3 on quarterly surface sample, monthly on ground water sample .
* IDM 13El    4.2 mi. w of  vent; Diehl House Lab      IDM 13E3    4.9 mi. w of  vent; local                VGT 16El    4.1 mi. NNW of vent; Port Penn            AIO, APT, IDM lFl    5.8 mi. N of vent; Fort Elfsborg          AIO, APT, IDM
16 15 14 12 ll FJGURE E-1 OFFSITE SAMPLING LOCATIONS 8 9 2 MANNINGTON 7 6 4 5
_lF2    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
.. 14 i ** 15 Cl::'.'. w > f-1 Cl::'.'. 11 GE FIGURE E-2 ONSITE SAMPLING LOCATIONS 1 3 5 M UM EXCLUSI AREA BOUNDA C901 METE ) R ER 6 7 10 N 9 J}}
 
Salem ODCM Rev. 6  3/28/90 TABLE E-1 (Cont'd)
STATION CODE                STATION LOCATION          SAMPLE TYPES 2F6    7.3 mi. NNE of vent; Southern Training  IDM Center 2F7    5.7 mi. NNE of vent; local farm          MLK, VGT 3F2    5 .1 mi. NE of vent; Hancocks Bridge    IDM Municipal Building 3F3    a. 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 7F2    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          IDM 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          MLK, VGT 14F2    6.6 mi. WNW of vent; Boyds Corner        IDM 14F3    5.4 mi. WNW of vent; local farm          FPV 15F3    5.4 mi. NW of vent                      IDM E-5
 
Salem ODCM Rev. 6   3/28/90 TABLE E-1 (Cont'd)
STATION CODE                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 lG3    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 Wilmington  IDM Airport 3Hl    32 mi. NE of vent; National Park, NJ    IDM 3H3    110 mi. NE of vent; Research and        AIO, APT, IDM Testing 3H5    25 mi. NE of vent; local farm          FPL, FPV E-6
 
Salem ODCM Rev. 6  3/28/90 SAMPLES COLLECTION AND ANALYSIS Sample          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 thoron 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 isotopic 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. 6    3/28/90
* SAMPLE COLLECTION AND ANALYSIS (Cont'd)
Sample          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 providing winter  monthly H-3 on Surface)        icing conditions allow. quarterly surface sample, monthly on ground water sample .
* E-8
 
FJGURE E-1 OFFSITE SAMPLING LOCATIONS 16 2
MANNINGTON 15 14 4
5 12 6
ll 7
8 9
~.
 
..                                   FIGURE E-2 ONSITE SAMPLING LOCATIONS 1
15 3
GE 14 Cl::'.'.
w
      >
f-1 Cl::'.'.
i                                                                5 M   UM EXCLUSI AREA BOUNDA C901 METE )
R ER   6 11 7
10                             N 9
**
J}}

Revision as of 11:51, 21 October 2019

Rev 6 to Odcm.
ML18095A447
Person / Time
Site: Salem  PSEG icon.png
Issue date: 04/11/1990
From: Palyz V
Public Service Enterprise Group
To:
Shared Package
ML18095A444 List:
References
PROC-900411, NUDOCS 9009070180
Download: ML18095A447 (98)
v • d • e


Text

Salem Generating Station Revision # 6 of the ODCM Revision Date: 4/11/90 Unit ~ Dock~t No. 50-272 Unit 2 Docket No. 50-311 Operating License No. DPR-70 Operating License No. DPR-75

  • .
    • 1 J

I

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PS~G

SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Revision 6 03/28/90 Approval SORC Chairman:~; Date:

Salem ODCM Rev. 6 03/28/90

1. 0 Liquid Effluents 1.1 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 ....... 5 1.3 Liquid Effluent Concentration Limits - 10 CFR 20 7 1.4 Liquid Effluent Dose Calculations - 10 CFR 50 . . . 8 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 2.0 Gaseous Effluents 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 . * . . . * . . 21 2.4 Noble Gas Effluent Dose Calculations - 10 CFR 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 . . . . 27 2.5.2 Simplified Dose Calculation for Radioiodines 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 . . . . 33 3.2.1 Effluent Dose Calculations ....... 35 3.2.2 Direct Exposure Determination ....... 35

Salem ODCM Rev. 6 03/28/90 Table of Contents - Continued 4.0 Radiological Environmental Monitoring Program .

  • 36 4.1 Sampling Program . . . . . . . . . . . . . .
  • 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, Aio . . . . . . . . . * . . . . . . . 43 1-4 Bioaccumulation Factors (BFi) * . . . . 45 2-1 Dose Factors for Noble Gases . . . * . . . . . . . 48 2-2 Parameters for Gaseous Alarm Setpoint Determinations

- Unit 1 . . . . . . . . . . . . . . . . . . . . . 49 2-3 Parameters for Gaseous Alarm Setpoint Determinati*ons

- Unit 2 . . . . . . . . . . . . . . . . . . . . . 5 0 2-4 Controlling Locations,* Pathways and Atmospheric Dispersion for Dose Calculations . . . . . . . 51

.- 2-5 A-1 A-2 B-1 B-2 Pathway Dose Parameters - Atmospheric Releases . . 52 Calcu~ation of Effective MPC - Unit 1 * . . . . . . A-4 Calculation of Effective MPC - Unit 2 . . . . . . . A-5 Adult Dose Contributions Fish and Drinking Water Pathways Unit 1 * * . . * . . . . . . . . . . .

Adult Dose Contributions Fish and Drinking Water

. B-5 Pathways Unit 2 * . * * . * . . * . . . . . . . . B-5 C-5 Effective Dose Factors . * * . . . . . . . . . C-5 Appendices Appendix A - Evaluation of Conservative, Default MPC Value for Liquid Effluents * * . * . . . A-1 Appendix B - Technical Basis for Effective Dose Factors -

Liquid Radioactive Effl~ents * . . . . . B-1 Appendtx C - Technical Bases for Effective Dose Factors -

Gaseous Radioactive Effluents . . . C-1 Appendix D - Radiological Environmental Monitoring Program -

Sample Type, Location and Analysis . . . D-1

Salem ODCM Rev. 6 03/28/90 SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL 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 50, 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 methodologies and parameters as is deemed necessary to ensure reasonable conservatism in keeping with the principles of 10 CFR 50.36a and Appendix I for demonstrating radioactive 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. 6 03/28/90

  • 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 (and Automatic Termination) R18 (Unit 1) and 2-R18 (Unit 2) provide the alarm and automatic termination of liquid radioactive material releases as required by Technical Specification 3.3.3.8.

1-R19 A,B,C,and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines.

2-R19 A,B,C and D provide this function for Unit 2 .

  • 2) Alarm Conly) - The alarm functions for the Service Water 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 1 and 2, respectively.

~-

2

-. 1.2 Salem ODCM Rev. 6 03/28/90 Liquid Effluent Monitor setpoint Determination Per the requirements of Technical Specification 3.3.3.8, alarm setpoints shall be established for the liquid eff~uent 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 Ir,* Column 2, for radionuclides and 2.0E-04 uCi/ml for dissolved or entrained noble gases). The following equation* must be satisfied to meet the liquid effluent restrictions:

C (F+f) (1.1) c ~

f where:

C = the effluent concentration limit of Technical 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, c< c. Also, note that when (F) is large compared to (f), then (F + f) = F.]

Adapted from NUREG-0133 3

Salem ODCM Rev. 6 03/28/90 1.2.1 Liquid Effluent Monitors (Radwaste, 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 ~ + bkg ( 1. 2)

RR with:

MPCe = ------------ ( 1. 3)

C*l.

~ --------

MPCi where:

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

MPCe = an effective MPC value for the mixture of radionuclides in the effluent stram (uCi/ml)

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

MPCi = the MPC value corresponding 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 (gal/min)

RR = the liquid effluent release rate (gal/min) bkg = the background of the monitor (cpm) 4

Salem ODCM Rev. 6 03/28/90 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.

lo2.2 Conservative Default Values. Conservative 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 the following:

a) substitution of the effective MPC value with a default value of lE-05 uCi/ml for radwaste releases (refer to Appendix A for justification) ;

b) for additional conservatism*, substitution of the I-131 MPC value of 3E-07 uCi/ml for the R19 Steam Generator

  • Use of the effective MPC value as derived in Appendix A may be non- conservative for the R19 Steam Generator blowdown monitors and R37 Chemical Waste Basin monitors where I-131 transfer during primary to secondary leakage may potentially be more controlling.blowdown monitors, 1R13** Service
  • Water monitor and R37 Chemical Waste Basin monitor; 5

Salem ODCM Rev. 6 03/28/90 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 MPCe and for differing Circulating Water System dilutions.

    • The Unit 2 Service Water system utilizes the Unit 1 Circulating Water system 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 2Rl3 default alarm setpoint. Because no dilution is considered and the 2Rl3 monitor sensitivity is high, the MPCe of lE-05 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. 6 03/28/90 1.3 Liquid Effluent concentration Limits - 10 CFR 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 limits of Technical Specification 3.11.1.1 may be performed using the following equation:

C*1 RR

~ ------ * ~ 1 (1.4)

MPC*1 ~+RR where:

C*1 = actual concentration of radionuclide i as measured in the undiluted liquid effluent (uCi/ml)

MPCi =-- 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. 6 03/28/90

  • 1.4 1.4.1 Liquid Effluent Dose Calculation - 10 CFR so MEMBER OF THE PUBLIC Dose - Liquid Effluents.

Technical Specification 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;

~ 1.5 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.

1. 67E-02
  • VOL
  • I: (C 1*
  • A*10 ) (1.5) cw where:

00 = dose or dose commitment to organ o, including total body (mrem)

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

Ci = 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. 6 03/28/90 The site-related ingestion dose/dose commitment factors (Aio> are presented in Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:

= 1.14E+05 [(UI

  • Bii) + (UF
  • BFi)J DFi ( 1. 6) where:

Aio = 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)

DFi = dose conversion factor for nuclide i for adults in pre-selected organ, o, from Table E-11 of Regulatory Guide 1.109 (mrem/pCi) l.14E+05 = conversion factor (pCi/uCi

  • ml/kg per hr/yr)

The radionuclides included in the periodic dose assessment per the requirements of Technical Specification 3/4.11.1.2 are those 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.

Radionuclides requiring radiochemical analysis (e.g., Sr-89 and Sr-90) will be added to the dose analysis at a frequency consistent with the required minimum analysis frequency of Technical Specification Table 4.11-1.

9

Salem ODCM Rev. 6 03/28/90

  • 1.4.2 simplified Liquid Effluent Dose Calculation.

the individual radionuclide dose assessment as presented in In lieu of 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

= ( 1. 7) cw

  • Maximum Organ
  • 2.52E+04 VOL

= * ( 1. 8) cw where:

C*l. = 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)

Dtb = conservatively evaluated total body dose (mrem)

Dmax = 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 Rev. 6 03/28/90 1.s Secondary Side Radioactive Liquid Effluents and Dose Calculations During 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 discharges.

Potentially significant radioactive material levels and potential releases are controlled/monitored by the Steam Generator blowdown monitors (R19) 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. 6 03/28/90

  • 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. 6 03/28/90 1.6 Liquid Effluent Dose Projections 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 individually. Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing requirements for the other unit.

Dose projections are made at least once per 31 days by the following equations:

Dtbp = Dtb (91 I d) ( 1. 9) 0 maxp = Dmax (91 I d) ( 1. 10) 13

Salem ODCM Rev. 6 03/28/90 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) 0 maxp = the maximum organ dose projection for current calendar quarter (mrem) 0 max = 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

Salem ODCM Rev. 6 03/28/90 2.0 Gaseous Effluents 2.1 Radiation 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. The tanks are discharged via the Plant Vent.

1-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 1-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-R12A and 2-R12A may be used to provide the containment monitoring and automatic isolation function during purge and pressure/vacuum reliefs.*

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 MdDE-6 .

15

Salem ODCM Rev. 6 03/28/90

3) Plant Vent (cont'd) R16) 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 rotometer).

A 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

  • Salem ODCM Rev. 6 03/28/90 2.2 Gaseous Effluent Monitor Setpoint Determination 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 release rate are used to calculate the fraction of the allowable release rate, as limited by Specification 3.11.2.1, by the equation:

FRAC = [4.72E+02

  • X/Q *VF* L (Ci* Ki)]/ 500 ( 2. 1)

FRAC = [4.72E+02

  • X/Q *VF* L (Ci* (Li+ 1.1 Mi))] / 3000 (2.2) where: --

FRAC = fraction of the allowable release rate based on the identified radionuclide concenrations and the release flow rate X/Q = annual average meteorological dispersion to the controlling site boundary location (sec/m 3 )

VF = ventilation system flow rate for the applicable release point and monitor (ft 3 /min)

C*l. = concentration of noble gas radionuclide i as determined by radioanalysis of grab sample (uCi/cm 3 )

K*l. = total body dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) 17

Salem ODCM Rev. 6 03/28/90

.

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

M*1 = gamma air dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad) 500 = total body dose rate limit (mrem/yr) 3000 = skin dose rate limit (mrem/vr) 4.72E+02 = conversion factor (cm3 /ft 3

  • min/sec)

Based on the more limiting FRAC (i.e., higher value) as determined above, the alarm setpoints for the applicable monitors (Rl6; R41C, and/or Rl2A) may be calculated by the equation:

SP = [AF * ~ Ci

  • SEN / FRAC] + bkg (2.3) where:

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

SEN = monitor sensitivity (cpm per uCi/cm 3 )

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 imposed to ensure that combined releases from Salem Units 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 0.5 (0.25 per unit), with the remainder o.s 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 18

Salem ODCM Rev. 6 03/28/90

  • allocation factors for all units do not exceed 1.0.

2.2.2 Conservative 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* 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 on 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 Specifications, Table 6 19

Salem ODCM Rev. 6 03/28/90 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations - 10 CFR 20 2.3.1 Site 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 5500 mrem/yr, total body and 53000 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 Ds = X/Q * (2.5)

' where:

Dtb = total body dose rate (mrem/yr)

Ds = skin dose rate (mrem/yr)

X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m 3 )

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

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

L*1 = beta skin dose conversion factor for noble gas radionuclide i (mrem/yr per uCi/m 3 , from Table 2-1)

M*1 = gamma air dose conversion factor for noble gas

  • radionuclide i (mrad/yr per uCi/m 3 , from Table 2-1) 1.1 = mrem skin dose per mrad gamma air dose (mrem/mrad)

As appropriate, simultaneous releases from Salem Units 1 and 2 and Hope Creek will be considered in evaluating compliance with the release rate limits of Specification 3.11.2.la, following any 20

  • Salem ODCM Rev. 6 03/28/90 release exceeding the above prescribed alarm setpoints. 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.

Actual meteorological conditions concurrent with the release

-

period or the default, annual average dispersion parameters as presented in- -Table 2-4 may be used for evaluating the gaseous effluent dose rate.

2.3.2 Site Boundary Dose Rate - Radioiodine and Particulates.

Technical Specification 3.11.2.1.b limits the dose rate to ~1500 mrem/yr to any organ for I-131, tritium and particulates with 21

Salem ODCM Rev. 6 03/28/90 half-lives greater than 8 days. To demonstrate compliance with this 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:

= X/Q * (2.6) where:

Do = 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/m31 Rio = dose parameter for radionuclide i (mrem/yr per uci/m )

and organ o for the child inhalation pathway from I

Table 2-5 Qi -. average release rate over the appropriate sampling period and analysis frequency for radionuclide i --

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

By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for I-131 can be determined. Based on the annual average meteorological dispersion (see Table 2-4) and the most limiting potential pathway, age group and organ (inhalation, child, thyroid -- Ri = 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 22

Salem ODCM Rev. 6 03/28/90

  • release rate allocated to each of the Salem 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.

23

Salem ODCM Rev. 6 03/28/90

  • 2.4 Noble Gas Effluent Dose Calculations - 10 CFR 50 2.4.1 UNRESTRICTED 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 dose limits of 55 mrad, gamma-air and 510 mrad, beta-air and the calendar year limits 510 mrad, gamma-air and 520 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
  • L (Mi
  • Qi) (2.7) and

= 3.17E-08

  • X/Q
  • L (Ni
  • Qi) (2.8)
  • where:

Dg

~

=

=

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

X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location (sec/m 3 )

Qi = cumulative release of noble gas radionuclide i over the period of interest (uCi)

Mi = air dose factor due to gamma emissions from noble gas radionuclide i (mrad/yr per uci;m 3 , from Table 2-1)

Ni = air dose factor due to beta emissions from noble gas radionuclide i (mrad/yr per uCi/m 3 , Table 2-1) 3.17E-08 = conversion factor (yr/sec)

    • 24

Salem ODCM Rev. 6 03/28/90 2.4.2 Simplified 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 3.11.2.2. (Refer to Appendix c for the derivation and justification for this simplified method.)

3.17E-08 Dg = --------

a.so

  • X/Q
  • Mef f
  • L Qi (2.9) and 3.17E-08 Db = --------
  • X/Q
  • Neff
  • L Qi (2.10) 0.50 where:

Mef f = 5.3E+02, effective gamma-air dose factor (mrad/yr per uCi/m 3 )

Neff = 1.1E+03, effective beta-air dose factor (mrad/yr per uCi/m 3 )

Qi = cumulative release for all noble gas radionuclides (uCi) 0.50 = conservatism factor to account for potential variability in the radionuclide distribution

.-

Actual meteorological conditions concurrent with 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. 6 03/28/90 2.5 Radioiodine and Particulate Dose Calculations - 10 CFR 50 2.5.1 UNRESTRICTED AREA Dose - Radioiodine and Particulates.

In accordance with requirements of Technical Specification 3.11.2.3, a periodic assessment shall be performed to evaluate compliance with the quarterly dose limit of 57.5 mrem and calendar year limit 515 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 a days:

0 aop = 3.17E-08

  • W
  • SFp * ~ (Riop
  • Qi) (2.11) where:

0 aop = 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 identified in Table 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)

= dose factor for radionuclide i (mrem/yr per uCi/m 3 )

or (m2 - mrem/yr per uCi/sec) and organ o from Table 2-5 for each age group a and the applicable pathway p as identified in Table 2-4. Values for Rio 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 greater than 8 days (uCi).

SFp = annual seasonal correction factor to account for the fraction of the year that the applicable exposur_e 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 Rev. 6 03/28/90

  • 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. 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).

0 max = 3.17E-08

  • W
  • SFp
  • RI-131 * ~ Qi ( 2 . 12)
  • where:

0 max RI-131

=

=

maximum organ dose (mrem)

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

= l.05E+12, infant thyroid dose parameter with the cow-milk pathway controlling (m2 - mrem/yr per uCi/sec) w = D/Q for radioiodine, 2.lE-10 1/m2 Qi = cumulative release over the period of interest for radionuclide i -- I-131 or radiaoctive material in particulate from with half life greater than 8 days (uCi)

The location of exposure pathways and the maximum organ dose 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

Salem ODCM Rev. 6 03/28/90 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. 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., R15 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., releases due to 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, the releases of the radioiodines- *and particulates shall be calculated by the equation:

= (2.13) where:

= the release rate of radionuclide, i, from the steam gener~tor flash tank vent (uCi/sec)

    • 28

Salem ODCM Rev. 6 03/28/90 C*1 = 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 SQftv = 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 (SQftv> 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 (pft) is approximately 0.48. The equation simplifies to the following:

(2.14)

For H-3, the simplified equation is:

(2.15)

Also during reactor shutdown operations with a radioactively contaminated secondary system, r~dioactive material may be released to the atmosphere via the atmospheric reliefs (PORV) and 29

Salem ODCM_ Rev. 6 03/28/90

  • the safety 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) I where:

= 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

= soo,ooo lb/hr per safety relief valve

= __ 50, ooo lb/hr for auxiliary feed pump exhaust PF*1 = partitioning factor, ratio of concentration in steam to that in the water in the steam generator

= 0.01 for radioiodine~

= 0.005 for all other particulates

= 1.0 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 of the Salem Emergency Plan Implementation Procedure .

    • 30

Salem ODCM Rev. 6 03/28/90 Alternately, the quantification of the release rate and cumulative releases may be based on actual samples of main steam collected at the R46 sample locations. The measured radionuclide concentration in the steam may be used for quantifying the noble gases, radioiodine and particulate releases.

Note: The expected mode of operation would be to isolate the effected steam generator, thereby reducing the potential releases during 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 demonstrating compliance with the Radiological Effluent Technical Specifications.

31

Salem ODCM Rev. 6 03/28/90

  • 2.7 Gaseous Effluent Dose Projection 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:

Dgp = Dg * (91 I d) (2.17)

~p = ~ * (91 I d) (2.18)

Dmaxp = Dmax * (91 I d) ( 2

  • 19) where: -

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

Dg = gamma air dose to date for current calendar quarter as d.e~~rmined by Equation 2.7 or 2.9 (mrem)

~p = 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)

.Dmaxp = maxi~um organ dose projection for current calendar quarter (mrem)

Dmax = 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

. Salem ODCM Rev. 6 03/28/90

  • 3.0 3.1 Special Dose Analyses Doses Due To Activities Inside the SITE BOUNDARY 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 assessment of radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY.

There is one location on Artificial Island that is accessible to MEMBERS OF THE PUBLIC for activities unrelated to PSE&G operational and support activities. This location is the Second

  • Sun (visitor's center) located near the contractors gate for the Salem Nuclear Generating Station.

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 and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per visit per year. The default value for the meteorological dispersion data as pre~ented in Table 2-3 may be used if current year meteorology is unavailable at the time of NRC reporting.

However, a follow-up evaluation shall be performed when the data becomes available.

33

. Salem ODCM Rev. 6 03/28/90 3.2 Total dose to MEMBERS OF THE PUBLIC - 40 CFR 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). 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 ~rom 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

  • Salem ODCM Rev. 6 03/28/90 3.2.1 Effluent Dose 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.

3.2.2 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 would yield potentially significant off-site doses (i.e., in excess of 1 mrem per year to a MEMBER OF THE PUBLIC), that would require detailed evaluation for demonstrating compliance with 40 CFR 190.

However, should a situation exist whereby the direct exposure contribution is potentially significant, on-site measurements, off-site measure~ents and/or calculation techniques will be used determination of dose for assessing 40 CFR 190 compliance .

  • for 35

Salem ODCM Rev. 6 03/28/90

  • 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 iri 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 environment.

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

  • NOTE: 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

Salem ODCM Rev. 6 03/28/90 4.2 Interlaboratory comparison Program Technical Specification 3.12.3 requires analyses be performed on 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 Program results will be provided in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 6.9.1.10.

37

  • RADIATION MONITOR QUID RELEASES UNIT 2 FIGURE 1-2 I FROM VOl. CONIROI. IRNK 12 REACTOR LEIDOIN HX CCXUVH DARIN TIN:

I VENT TO GASEOUS J"

TO VENT TO VENT nlXED BED OEH I NERfl. I ZEA

- JO HDLD-ll' IRllCS COll'OIENT COil.iNG SlflllE-TRllC REFUELING IATER STORAGE to-JANI(

12 SPENT FIA:l PIT DEHINERIUER 121 HDLD-ll' TIN: J 122 HDLD-lJ' JIN( LJ 123 HDLD-ll' lfH:

VCUK IRSTE POU TOA CON!Rll. TIN: - Tfl<< Ill SRll'l.1116 SYSTEH l

-

STEii! GENElllTDR IRSTE HONITOA HDLD-IP TfNC 12 J' IRSTE HDLD-11' TRllC 121 r r lllSTE IO.IHI' TRllC m NO. 2 lllSTE EVll'ORATDR f--o lllSTE HONITDR TRllC 122 121 EVll'IJllTDR FEED ION EXClfNlER -- 124 EVll'ORATOA FEED ION EXClllNGER 122 EVll'OfWITOR FEED ION EXCllNlER I+---

l 123 EVll'ORATOR FEED ION EXClflNGER SAlf'\.E Llll:S A19 HON ITCJISITJ STEii!

.. STEii! NO. 2 6flS GE NEiii TORS BENElllTDR STRIPPER SALEM OOCM 121.22.23.24 STEAlt GENElllTDR BllllllOIN BLOIOOIN TRllC LINES <FDR 11.-CJRMA TI ON ONLY l

[!]RIB NON-APO LIQUID IASTE DISPOSAL All r1 TO CIRCutATING IATER SYSTEH

__J DELRIARE AI VER

RADIATION MON G LIQUID RELEASES UNIT 1

'

I fROl1 VOL. CIJN!AOl lnNK

..

LETOOIN HK II RERCTOR COOLllHT ORRIN TllNC

VENT TO GASEllJS "IXED BED DE"INEMLIZER

- - + TO HDLD-lJ' TllNCS CIJllOENT CIXLIHB SlllGE-TlllC REFl£LIN6 IATER STORAGE TllNC lo--

I I SPENT Fl£L PIT DE"INEMLZER II I HOLD-II' TllNC 1'T' 112 HOLD-lJ' TANK LJ~

llJ HDLD-lJ' TllNC J"

i VDLIK lllSTE 1111 ITlll CONTRll. TllNC ~

TANC Ill Slff'LIHB SYSTE" ~

lllSTE lllllTlll 111.D-lJ' TIN:

fl J- lllSTE Ill.IHI' TIN: Ill j- RITE Ill.IHI' TIN: Ill?

J- HO. I lllSTE EVll'lllATlll

...__. "'fTE lllllTlll IN: 112 E~Tlll FEED ION EXDINER 114 EVll'CRITlll FEED ION EXClllH6ER 112 EVll'lllATlll FEED lllf EXClllH6ER lo---

I llJ EYll'lllATIJI FEED lllf EKClflHGER l

-

R19 ITlllSITJ STEii! GENERATlll Sfff'LE LINES STElll GENEMTlllS 111.12.13,14 STElll GENEMTlll ILOIOOIN LINES STElll GENERllTlll ILOIOOIN TllNC HD* I &AS STRIPPER

(!] RIB NON-RAD LIOOID IASTE DISPOSll.

RJI ri TO CIRCl.lATIN6 IATER SYSTE"

_J SALEH OOCH If OR INflllltAI ION ONL YI DELRIFff RIVER

  • N vaporator package and/or r.adwaste demineralizer system LAUNDRY AND SllOWEa DllAINS LA* EQUIP. AND f'LOOA DaAINS LA8 DRAINS TO WASTE GAS COMPRESSOR IWL99 JWL91 LAUNDRY LAUNDRY CHE*llCAL GAS AND ANO DRAIN HOT ANALYZ£k HOT TANK SHOW EA IHOWl.R aADWASTI: ACCUMULATOll UllJ\INS PACKAGE Ht:ttJl:lJNG EXCESS LETlJOW N CAN:\L

' ltlACTOlt t"LANGl. U:At..Ut t*

SOLID RADIOACTIVE RWST LOUP UKAINS WASTll S\'STt:M eves (HIC)

  • .z

..< l'HESSUKllt.:H HJ::LU:r TANK 0

.

Q

....

0 IWL11

..z Q

., IWLl6 REACTOR COOLANT le z DRAIN TANK PUMPS

<

." ".

lE c

... ..

ci z tROM CONTAINJi.IENT

  • "

WASTE SUMP l'UMl'S

~

t.IONl10R TANK

~IONITUK TANK EVAP.'

FEED PU*IP . 5

"< C>

"'

\\ .\S"l t :0.101\:ITUH

  • rANK t"l!Ml'S Fig 1-3 Salem L111uicJ ltadioaclive W~h' ~y:-.lrm 1U IWL51 FOR INFORMATION ONLY l:IHC'lll.,\l"IN(i

~A'l"IJI ~\*~Tl:\;

'""' l.~bh

Salem ODCM Rev. 6 03/28/90 Table 1-1 Parameters for Liquid Alarm setpoint Determinations Unit 1 Parameter Actual Default Units Conments Value Value MPCe calculated 1E-05

  • uCi/ml calculated for each batch to be released MPCI-131 3E-07 N/A uCi/ml I-131 MPC conservatively used for SG blow-down and Service Water monitor setpoints Ci measured N/A uCi/ml taken from ganma spectral analysis of liquid effluent MPCi as N/A uCi/ml taken from 10 CFR 20, Appendix B, Table II, determined Col. 2.

SEN 1-R18 as 2.9E+07 cpm per uCi/ml radwaste effluent (Cs-137) determined 1-R19 2.9E+07 Steam Generator blowdown (Cs-137)

(A,B,C,D) 1-R13 1.2E+08 Service Water - Containment fan cooling (A,B,C,D,E) (Cs-137) cw as 1.85E+OS gpm Circulating Water System, single CW ~

determined RR 1-R18 as 120 gpm determined prior to release; release rate determined can be adjusted for Technical Specification compliance 1-R19 120 Steam Generator blowdown rate per generator 1-R13 2500 Service Water flow rate for Containment fan coolers SP 1-R18

  • calculated 4.4E+05(+bkg) cpm Default alarm setpoints; more conservative values may be used as deemed appropriate and 1-R19** calculated 1.3E+04C+bkg) desirable for ensuring regulatory compliance and for maintaining releases ALARA.

1-R13** calculated 2.6E+03(+bkg)

  • Refer to Appendix A for derivation

41

Salem ODCM Rev. 6 03/28/90 Table 1-2 Parameters for Liquid Alarm Setpoint Determinations Unit 2 Parameter Actual Default Units Conments Value Value MP Ce calculated 1E-05

  • uCi/ml calculate for each batch to be released MPCl-131 3E-07 N/A uCi/ml 1-131 MPC conservatively used for SG blow-down, Service Water and Chemical Waste Basin monitor setpoints Ci measured N/A uCi/ml taken from ganma spectral analysis of liquid effluent MPCi as N/A uCi/ml taken from 10 CFR 20, Appendix B, Table II, determined Col. 2.

SEN 2-R18 as 8.8E+07 cpm per uCi/ml radwaste effluent (Cs-137) determined 2-R19 8.8E+07 Steam Generator blowdown (Cs-137)

CA,B,C,D) 2-R13 8.8E+07 Service Water - Containment fan cooling CA,B,C) (Cs-137)

R37 8.8E+07 Chemical Waste Basin (Cs-137) cw as 1.85E+05 gpm Circulating Water System, single CW P\.11'>

determined (Note no CW P\.11'> in service for 2R13 monitor see section 1.2.2)

RR 2-R18 as 120 gpm determined prior to release; release rate determined can be adjusted for Technical Specification compliance 2-R19 120 Steam Generator blowdown rate per generator 2-R13 2500 Service water flow rate for Containment fan coolers R37 300 Chemical Waste Basin discharge SP 2-R18** calculated 8.0E+OSC+bkg) cpm Default alarm setpoints; more conservative values may be used as deemed appropriate and 2-R19*** calculated 3.9E+04C+bkg) desirable for ensuring regulatory compliance and for maintaining releases ALARA.

2-R13*** calculated 8.8E+02C+bkg)

R37**** calculated 1_.6E+04C+bkg)

  • Refer to Appendix A for derivation
    • Actual calculated setpoint for 2-R18 (1.3E+06) is greater than the full scale monitor indicator, therefore, for conesrvatism the recommended-setpoint has been reduced to 3.0E+OS cpm
        • The MPC value of 1-131 (3E-07 uCi/ml) has been used for derivation of the R19 Steam generator blowdown and the R37 Chemical Waste Basin monitor setpofnts as discussed in Section 1.2.2

Salem ODCM Rev. 6 03/28/90 Table 1-3 Site Related Ingestion Dose Commitment Factors, Aio (mrem/hr per uci/ml)

Nuclide Bone Liver T.Body Thyroid Kidney

------- -------- ------- ------- ------- -------

Lung _____

GI-LLI

_.._

H-3 2.82E-1 2.82E-1 2.82E-1 2.82E-1 2.82E-l 2.82E-l C-14 1. 45E+4 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 2.90E+3 Na-24 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-1 4.57E-l 4.57E-l P-32 4.69E+6 2.91E+5 1.81E+5 5.27E+5 Cr-51 5.58E+O 3.34E+O 1.23E+O 7.40E+O 1. 40E+3 Mn-54 7.06E+3 1. 35E+3 2.10E+3 2.16E+4 Mn-56 1. 78E+2 3.15E+l 2.26E+2 5.67E+3 Fe-55 5.11E+4 3.53E+4 8.23E+3 1.97E+4 2.03E+4 Fe-59 8.06E+4 1. 90E+5 7.27E+4 5.30E+4 6.32E+5 Co-57 1.42E+2 2.36E+2 3.59E+3 Co-58 6.03E+2 1. 35E+3 l.22E+4 Co-60 1. 73E+3 3.82E+3 3.25E+4 Ni-63 4.96E+4 3.44E+3 1. 67E+3 7.18E+2 Ni-65 2.02E+2 2.62E+l 1. 20E+l 6.65E+2 cu-64 2.14E+2 1. 01E+2 5.40E+2 1. 83E+4 Zn-65 1. 61E+5 5.13E+5 2.32E+5 3.43E+5 3.23E+5 Zn-69 3.43E+2 6.56E+2 4.56E+l 4.26E+2 9.85E+l Br-82 4.07E+O 4.67E+O Br-83 7.25E-2 l.04E-l Br-84 9.39E-2 7.37E-7 Br-85 3.86E-3 Rb-86 6.24E+2 2.91E+2 l.23E+2 Rb-88 1.79E+O 9.49E-1 2.47E-ll Rb-89 l.19E+O 8.34E-1 6.89E-14 Sr-89 4.99E+3 1. 43E+2 8.00E+2 Sr-90 1. 23E+5 3.01E+4 3.55E+3 Sr-91 9.18E+l 3.71E+O 4.37E+2 Sr-92 3.48E+l 1. 51E+O 6.90E+2 Y-90 6.06E+O 1. 63E-l 6.42E+4 Y-91m 5.73E-2 2.22E-3 1.68E-l Y-91 8.88E+l 2.37E+O 4.89E+4*

Y-92 5.32E-1 1. 56E-2 9.32E+3 Y-93 l.69E+O 4.66E-2 5.35E+4 Zr-95 1. 59E+l 5.llE+O 3.46_E+O 8.02E+O l.62E+4 Zr-97 8.81E-1 1.78E-1 8.13E-2 2.68E-1 5.51E+4 Nb-95 4.47E+2 2.49E+2 1. 34E+2 2.46E+2 1.51E+6 Nb-97 3.75E+O 9.49E-1 3.46E-1 1.llE+O 3.50E+3

,, .Mo-99 1.2SE+2. 2.43E+l 2.89E+2 2.96E+2 Tc-99m 1.30E-2 3.66E-2 4.66E-1 5.56E-l 1.79E-2 2.17E+l Tc-101 1. 33E-2 1. 92E-2 1. 88E-1 3.46E-1 9.81E-3 5.77E-14 43

Salem ODCM Rev. 6 03/28/90

  • 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 3.51E+O l.15E+2 5.44E+3 Ru-106 1. 59E+3 2.01E+2 3.06E+3 1.03E+5 Rh-103m Rh-106 Ag-llOm 1.56E+3 1.45E+3 8.60E+2 2.85E+3 5.91E+5 Sb-124 2.77E+2 5.23E+O 1.10E+2 6.71E-1 2.15E+2 7.86E+3 Sb-125 1.77E+2 1. 98E+O 4.21E+l 1. 80E-1 1.36E+2 1. 95E+3 Te-125m 2.17E+2 7.86E+1 2.91E+l 6.52E+l 8.82E+2 8.66E+2 Te-127m 5.48E+2 1. 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 1.47E+2 3.20E+2 3.89E+3 4.69E+3 Te-129 2.54E+O 9.55E-1 6.19E-1 1. 95E+O 1. 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 1.17E+2 4.61E+l 9.91E+3 1.82E+2 1.01E+2 I-131 2.18E+2 3.12E+2 1.79E+2 1. 02E+5 5.35E+2 8.23E+l I-132 1.06E+l 2.85E+l 9.96E+O 9.96E+2 4.54E+l 5.35E+O I-133 7.45E+l 1. 30E+2 3.95E+l 1. 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. 3-2E+l 6.08E+l 2.24E+l 4.01E+3 9.75E+l 6.87E+l Cs-134 ~.84E+3 1. 63E+4 1. 33E+4 5.27E+3 1.75E+3 2.85E+2 Cs-136 7.16E+2 2.83E+3 2.04E+3 1.57E+3 2.16E+2 3.21E+2 Cs-137 8.77E+3 1. 20E+4 7.85E+3 4.07E+3 1.35E+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 7.85E+O 5.59E-3 2.30E-1 5.23E-3 3.17E-3 1. 39E+l Ba-140 1.64E+3 2.06E+O 1.08E+2 7.02E-1 1.18E+O 3.38E+3 Ba-141 3.81E+O 2.88E-3 1.29E-1 2.68E-3 1.63E-3 1.SOE-9 Ba-142 1.72E+O 1.77E-3 1. OSE-1 1. SOE-3 1.00E-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 3.43E+O 2.32E+O 2.63E-1 1.0SE+O 8.86E+3 Ce-143 6.04E-1 4.46E+2 4.94E-2 1.97E-1 1.67E+4 Ce-144 1.79E+2 7.47E+l 9.59E+O 4.43E+l 6.04E+4 Pr-143 5.79E+O 2.32E+O 2.87E-1 1. 34E+O 2.54E+4 Pr-144 1. 90E-2 7.87E-3 9.64E-4 4.44E-3 2.73E-9 Nd-147 3.96E+O 4.58E+O 2.74E-1 2.68E+O 2.20E+4 W-187 9.16E+O 7.66E+O 2.68E+O 2.51E+3 Np-239 3.53E-2 3.47E-3 1. 91E-3 1.08E-2 7.11E+2 44

Salem ODCM Rev. 6 03/28/90

  • Table 1-4 Bioaccumulation Factors (BFi)

(pCi/kg per pCi/liter)*

Element Saltwater Fish Saltwater Invertebrate H 9.0E-01 9.3E-Ol c l.8E+03 1. 4E+03 Na 6.7E-02 1.9E-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 l.OE+02 l.OE+03 Ni 1.0E+02 2.5E+02 Cu 6.7E+02 l.7E+03 Zn 2.0E+03 5.0E+04 Br 1.5E-02 3.lE+OO Rb 8.3E+OO l.7E+Ol Sr 2.0E+OO 2.0E+Ol y 2.5E+Ol l.OE+03 Zr 2.0E+02 8.0E+Ol Nb 3.0E+04 l.OE+02 Mo 1. OE+Ol l.OE+Ol Tc 1. OE+Ol 5.0E+Ol Ru 3.0E+OO 1. OE+03 Rh 1. OE+Ol 2.0E+03 Ag 3.3E+03 3.3E+03 Sb 4.0E+Ol 5.4E+OO Te 1. OE+Ol l.OE+02 I l.OE+Ol 5.0E+Ol Cs 4.0E+Ol 2.5E+Ol Ba 1. OE+Ol l.OE+Ol La 2.5E+Ol 1. OE+03 Ce l.OE+Ol 6.0E+02 Pr 2.5E+Ol l.OE+03 Nd 2.5E+Ol l.OE+03 w 3.0E+Ol 3.0E+Ol Np 1. OE+Ol l.OE+Ol

~- 45

  • ~

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Salem ODCM Rev. 6 03/28/90

  • Table 2-1 Dose Factors for Noble Gases Total Body Galllll8 Air Beta Air Dose Factor Skin Dose Factor Dose Factor Dose Factor Radionuclide Ki Li Mi Ni Cmrem/yr per uCi/m3) Cmrem/yr per uCi/m3) Cmrad/yr per uCi/m3) Cmrad/yr per uCi/m3)

-------------------- -------------------- ------------------- --------------------

Kr-83m 7.56E-02 1.93E+01 2.88E+02 Kr*85m , .17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1.nE+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6. 17E+03 1.03E+04 Kr-88 1 .47E+04 2.37E+03 1.52E+04 2.93E+03 Kr-89 1.66E+04 1.01E+04 1. 73E+04 1.06E+04 Kr-90 1.56E+04 7.29E+03 1.63E+04 7.83E+03 Xe* 131m 9. 15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-133m 2.51E+02 9.94E+02 3.27E+02 1.48E+03 Xe-133 2.94E+02 3.06E+02 3.53E+02 1.05E+03 Xe-135m 3. 12E+03 7. 11E+02 3.36E+03 7.39E+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. 6 03/28/90

  • Table 2-2 Parameters for Gaseous Alarm Setpoint Determinations Unit-1 Parameter Actual Default Units Cooments Value Value

-------- -------- ---------- -------------------------------------------

X/Q calculated 2.2E-06 sec/m3 USNRC Salem Safety Evaluation, Sup. 3 VF as measured 1.25E+05 ft 3/min Plant Vent - normal operation (Plant or Vent) fan curves VF as measured 3.5E+04 Containment purge (Cont. or Purge) fan curves AF coordinated 0.25 unit less Acministrative allocation factor to.*

with HCGS ensure conbined re.Leases do not exceed release rate Limit for site.

Ci measured N/A uCi/cm3 Ki nuclide specific*- N/A mrem/yr per uCi/m3 Values from Table 2-t Li nuclide specific N/A mrem/yr per uCi/m3 Values from Table 2-1 Mi nuclide specific N/A mrad/yr per uCi/m3 Values from Table 2-1 SEN 1-R41C* as 1.6E+07 cpm per uCi/cm3 Plant Vent determined 1-R16 3.6E+07 Plant Vent (redundant)

-

1-R12A 2.1E+06 Containment SP 1-R41C calculated 3.3E+04C+blcg) cpm Default alarm setpoints; more conservative values may be used as deemed appropriate and 1-R16 calculated 7.4E+04C+blcg) desirable for ensuring regulatory coq:>liance and for maintaining releases ALARA.

1-R12A** calculated 1.5E+04(+blcg)

  • Based on mean.for calibration with mixture of radionuclides
    • Applicable during MODES 1 through 5. During MODE 6 (refueling), monitor setpoint shall be reduced to 2X baclcgrourid in accordance with Tech Spec Table 3.3-6.

49

Salem ODCM Rev. 6 03/28/90 Table 2*3 Parameters for Gaseous Alarm Setpoint Determinations Unit-2 Parameter Actual Default Units Comments Value Value

-------- -------------------------------------------

X/Q calculated 2.2E-06 sectm3 Licensing technical specification value*

VF as measured 1.25E+OS Plant Vent - normal operation CPL ant or Vent) fan curves VF as measured 3.SE+04 Containment purge (Cont. or Purge) fan curves AF coordinated 0.25 unitless Administrative allocation factor to with HCGS ensure coat>ined releases do not exceed release rate Limit for site.

Ci measured N/A uCi/cm3

  • Ki Li Mi nuclide specific nuclide specific nuclide specific N/A N/A N/A mrem/yr per UCitm3 mrem/yr per uCitm3 mrad/yr per uCitm3 Values from Table 2-1 Values from Table 2-1 Values from Table 2-1 SEN 2-R41C* as 1.6E+07 cpm per uCi/cm3 Plant Vent determined 2-R16 3.SE+07 Plant Vent (redundant) 2-R12A 3.3E+07 Containment SP 2-R41C calculated 3.3E+04C+bkg) cpm Default alarm setpoints; more conservative values may be used as deemed appropriate and 2-R16 calculated 7.2E+04(+bkg) desirable for ensuring regulatory c~Liance and for maintaining releases ALARA.

2-R12A** calculated 2.4E+OSC+bkg)

  • Based on mean for calibration with mixture of radionuclides
    • Applicable during MOOES 1 through 5. During MOOE 6 (refueling), monitor setpoints shall be reduced to 2X background in accordance with Tech Spec Table 3.3-6.

50

Salem ODCM Rev. 6 03/28/90 Table 2-4 Controlling Locations, Pathways and Atmospheric Dispersion for Dose Calculations

  • Atmospheric Dispersion Technical Specification Location Pathway(s) Controlling X/Q D/Q Age Group (sec/m3 > C1tm2 >

3.11.2.1a site boundary noble gases N/A 2.2E-06 N/A (0.83 mi le, N) direct exposure 3.11.2.1b site boundary inhalation child 2.2E-06 N/A (0.83 mile, N) 3.11.2.2 site boundary gamma-air N/A 2.2E-06 N/A (0.83 mi le, N) beta-air 3.11.2.3 residence/dairy C4.9 miles, W) milk, ground plane and inhalation infant S.4E-08 2.1E-10 I

6.9.1.10 Second sun direct exposure N/A 8.22E-06 N/A

(0.21 mile/SE) and inhalation

  • The identified controlling locations, pathways and atmospheric dispersion are from the Safety Evaluation Report, Supplement No. 3 for the Salem Nuclear Generating Station; Unit 2 CNUREG-0517, December 1978).

51

-

Salem ODCM Rev. 6 03/28/90

  • 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 1.18E+4 2.78E+4 - - 1. 02E+6 l.88E+5 l.06E+4 Co-57 - 6.92E+2 - - 3.70E+5 3.14E+4 6.71E+2 Co-58 - 1.58E+3 - - 9.28E+5 1. 06E+5 2.07E+3 Co-60 - 1.15E+4 - - 5.97E+6 2.85E+5 1. 48E+4 Ni-63 4.32E+5 3.14E+4 - - 1.78E+5 l.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 - - - 1.66E+4 5.90E+4 Sr-89 3.04E+5 - - - 1. 40E+6 3.50E+5 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 1. 07E+5 3.44E+4 - 5.42E+4 1.77E+6 1. 50E+5 2.33E+4 Nb-95 1. 41E+4 7.82E+3 - 7.74E+3 5.05E+5 l.04E+5 4.21E+3 Ru-103 1.53E+3 - - 5.83E+3 5.05E+5 1.10E+5 6.58E+2 Ru-106 6.91E+4 - - 1. 34E+5 9.36E+6 9.12E+5 8.72E+3 Ag-llOm l.08E+4 1.00E+4 - 1. 97E+4 4.63E+6 3.02E+5 5.94E+3 Sb-124 - 3.12E+4 5.89E+2 7.55E+l - 2.48E+6 4.06E+5 1. 24E+4 Sb-125 Te-125m 5.34E+4 3.42E+3 5.95E+2 1.58E+3 5.40E+l 1.05E+3

- 1.74E+6 1. 01E+5 1.26E+4 4.67E+2 1.24E+4 3.14E+5 7.06E+4 Te-127m 1.26E+4 5.77E+3 3.29E+3 4.58E+4 9.60E+5 1.50E+5 1.57E+3 Te-129m 9.76E+3 4.67E+3 3.44E+3 3.66E+4 1.16E+6 3.83E+5 l.58E+3 I-131 2.52E+4 3.58E+4 1.19E+7 6.13E+4 - 6.28E+3 2.05E+4 Cs-134 3.73E+5 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 l.10E+5 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 1.35E+4 - 6.26E+3 3.62E+5 1.20E+5 1.53E+3 Ce-144 3.43E+6 1.43E+6 - 8.48E+5 7.78E+6 8.16E+5 1.84E+5 Pr-143 9.36E+3 3.75E+3 - 2.16E+3 2.81E+5 2.00E+S 4.64E+2

".-Nd-147 5.27E+3 6.lOE+3 - 3.56E+3 2.21E+5 1. 73E+5 3.65E+2 52

..

Salem ODCM Rev. 6 03/28/90 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 1. 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 1. 35E+2 Mn-54 5.11E+4 1.27E+4 1. 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 1.59E+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 5.80E+5 4.34E+4 3.07E+5 1.42E+4 1. 98E+4 Zn-65 3.86E+4 1.34E+5 8.64E+4 1.24E+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. 08E+8 l.65E+7 7.65E+5 6.68E+6 Y-91 6.61E+5 2.94E+6 4.09E+5 1.77E+4 Zr-95 1.46E+5 4.58E+4 6.74E+4 2.69E+6 1.49E+5 3.15E+4 Nb-95 1.86E+4 1. 03E+4 1.00E+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 l.61E+7 9.60E+5 1. 24E+4 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 1.59E-+5 2.18E+3 Te-129m l.39E+4 6.58E+3 4.58E+3 5.19E+4 1. 98E+6 4.05E+5 2.25E+3 I-131 3.54E+4 4_. 91E+4 1. 46E+7 8.40E+4 6.49E+3 2.64E+4 Cs-134 5.02E+5 1.13E+6 3.75E+5 1. 46E+5 9.76E+3 5.49E+5 Cs-136 5.15E+4 1.94E+5 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 2.28E+l 2.03E+6 2.29E+5 3.52E+3 Ce-141 2.84E+4 1.90E+4 8.88E+3 6.14E+5 1.26E+5 2.17E+3 Ce-144 4.89E+6 2.02E+6 1. 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. 6 03/28/90 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 1.12E+3

  • 1.12E+3 1.12E+3 1.12E+3 1.12E+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 1.14E+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 l.11E+5 2.87E+3 7.77E+3 Fe-59 2.07E+4 3.34E+4 l.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 1.13E+5 7.14E+4 9.95E+5 l.63E+4 7.03E+4*

Rb-86 1.98E+5 7.99E+3 1.14E+5 Sr-89 5.99E+5 2.16E+6 1.67E+5 1. 72E+4

  • sr-90 1. 01E+8 1. 48E+7 3.43E+5 6.44E+6 Y-91 9.14E+5 2.63E+6 1. 84E+5 2.44E+4 Zr-95 1. 90E+5 4.18E+4 5.96E+4 2.23E+6 6.11E+4 3.70E+4*

Nb-95 2.35E+4 9.18E+3 8.62E+3 6.14E+5 3.70E+4 6.55E+3 Ru-103 2.79E+3 7.03E+3 6.62E+5 4.48E+4 1. 07E+3 Ru-106 1.36E+5 1. 84E+5 1. 43E+7 4.29E+5 1.69E+4 Ag-llOm 1.69E+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 l.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 5.03E+4 1. 76E+6 l.82E+5 3.04E+3 I-131 4.81E+4 4.81E+4 l.62E+7 7.88E+4 2.84E+3 2.73E+4 Cs-134 6.51E+5 1. 01E+6 3.30E+5 1. 21E+5 3.85E+3 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 1. 04E+5 3.62E+3 1.28E+5 Ba-140 7.40E+4 6.48E+l 2.llE+l 1. 74E+6 l.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 l.17E+6 1.20E+7 3.89E+5 3.61E+5 Pr-143 l.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+S 8.21E+4 6.81E+2 I

.

I I

54

Salem ODCM Rev. 6 03/28/90 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

------- ------- ------- ------- ------- ------- ------- 6.47E+2 H-3 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 l.61E+4 7.74E+4 Cr-51 5.75E+l l.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 l.97E+4 1. l 7E+4 8.69E+4 l.09E+3 3.33E+3 Fe-59 l.36E+4 2.35E+4 1.02E+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 l.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 l.16E+4 Zn-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 l.14E+4

  • Sr-90 4.09E+7 l.12E+7 l.31E+5 2.59E+6 Y-91 5.88E+5 2.45E+6 7.03E+4 l.57E+4 Zr-95 l.15E+5 2.79E+4 3.11E+4 1.75E+6

.

2.17E+4 2.03E+4 Nb-95 l.57E+4 6.43E+3 4.72E+3 4.79E+5 1.27E+4 3.78E+3 Ru-103 2.02E+3 4.24E+3 5.52E+5 1.61E+4 6.79E+2 Ru-106 8.68E+4 l.07E+5 1.16E+7 1.64E+5 l.09E+4 Ag-llOm 9.98E+3 7.22E+3 l.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 l.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 l.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 -- 3.79E+4 4.44E+4 1.48E+7 5.18E+4 l.06E+3 1. 96E+4 Cs-134 3.96E+5 7.03E+5 1. 90E+5 7.97E+4 1.33E+3 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+5 7.13E+4 l.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 l.67E+4 5.25E+3 5.17E+5 2.16E+4 l.99E+3 Ce-144 3.19E+6 l.21E+6 5.38E+5 9.84E+6 1. 48E+5 1. 76E+5 Pr-143 1.40E+4 5.24E+3 l.97E+3 4.33E+5 3.72E+4 6.99E+2

.. Nd-147 7.94E+3 8.13E+3 3.15E+3 3.22E+5 3.12E+4 5.00E+2 55

Salem ODCM Rev. 6 03/28/90 Table 2-5 (cont'd)

R ( io) , Grass-Cow-Milk 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 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 l.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 1.60E+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 l.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 l.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+l0" 1. 35E+9 l.15E+10 Y-91 8.60E+3 4.73E+6 2.30E+2 Zr-95 9.46E+2 3.03E+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 l.32E+6 2.58E+3 Ag-llOm Sb-124 5.83E+7 2.57E+7 5.39E+7 4.86E+5

-

6.24E+4 1.06E+8 2.00E+7 2.20E+10 7.31E+8 3.20E+7 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 l.64E+7 1.17E+7 1. 86E+8 1.54E+8 5.58E+6 Te-129m 6.04E+7 2.25E+7 2.08E+7 2.52E+8 3.04E+8 9.57E+6 I-131 2 *. 96E+8 4.24E+8 1. 39E+ll 7.27E+8 l.12E+8 2.43E+8 Cs-134 5.65E+9 1.34E+10 4.35E+9 l.44E+9 2.35E+8 1.lOE+lO Cs-136 2.61E+8 l.03E+9 5.74E+8 7.87E+7 l.17E+8 7.42E+8 Cs-137 7.38E+9 1. OlE+lO 3.43E+9 1.14E+9 1.95E+8 6.61E+9 Ba-140 2.69E+7 3.38E+4 l.15E+4 1.93E+4 5.54E+7 1. 76E+6 Ce-141 4~84E+3 3.27E+3 1.52E+3 1. 25E+7 3.71E+2 Ce-144- 3.58E+5 1.50E+5 8.87E+4 1. 21E+8 1.92E+4 Pr-143 l.59E+2 6.37E+l 3.68E+l 6.96E+5 7.88E+O Nd-147 9.42E+l 1.09E+2 6.37E+l 5.23E+5 6.52E+O 56

Salem ODCM Rev. 6 03/28/90 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 l.34E+5 1. 34E+5 l.34E+5 P-32 3.15E+10 1. 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.i6E+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 l.10E+8 l.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 Y-91 Zr-95 Nb-95 Ru-103 6.61E+10 1.58E+4
1. 65E+3
1. 41E+5
1. 81E+3 5.22E+2 7.80E+4 7.67E+2 7.57E+4 6.40E+3 1.86E+9 6.48E+6 1.20E+6 3.34E+8 1.52E+5 1.63E+l0 4.24E+2 3.59E+2 4.30E+4 7.75E+2 Ru-106 3.75E+4 7.23E+4 1.80E+6 4.73E+3 Ag-110m 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 2.10E+8 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+11 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 l.07E+l0 Cs-136 4.45E+8 1. 75E+9 9.53E+8 1.50E+8 1.41E+8 l.18E+9.

Cs-137 1.34E+10 1.78E+10 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 2.92E+2 1.17E+2 6.77E+1 9.61E+5 1.45E+l

.

  • Nd-147 1. 81E+2 1. 97E+2
  • 1.16E+2 7.11E+5 1.18E+l 57

Salem ODCM Rev. 6 03/28/90

.

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+5 3.29E+5 P-32 7.77E+l0 3.64E+9 2.15E+9 3.00E+9 Cr-51 5.66E+4 1.55E+4 1. 03E+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 l.84E+7 Fe-59 1. 20E+8 1.95E+8 5.65E+7 2.03E+8 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 l.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 1.89E+8 Sr-90 1.12E+ll 1.51E+9 2.83E+l0 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+S 1.16E+5 2.29E+8 8.84E+4 Ru-103 4.29E+3 1.08E+4 1.llE+S 1. 65E+3 Ru-106 9.24E+4 1. 25E+5 1.44E+6 l.15E+4 Ag-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 l.41E+6 8.06E+4 4.85E+7 2.08E+8 1. 82E+7 Te-125m 7.38E+7 2.00E+7 2. 07E+7 *

  • 7.12E+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+S 3.32E+8 4.23E+7 I-131 l.30E+9 1. 31E+9 4.34E+ll 2.15E+9 1.17E+8 7.46E+8 cs-134 2.26E+l0 3.71E+10 l.15E+10 4.13E+9 2.00E+8 7.83E+9 cs-136 1.00E+9 2.76E+9 1. 47E+9 2.19E+8 9.70E+7 1.79E+9 Cs-137 3.22E+10 3.09E+10 1. OlE+lO 3.62E+9 1.93E+8 4.55E+9 Ba-140 l.17E+8 1.03E+5 3.34E+4 6.12E+4 5.94E+7 6.84E+6 Ce-141 2.19E+4 l.09E+4 4.78E+3 1.36E+7 1. 62E+3 Ce-144 l.62E+6 5.09E+S 2.82E+5 1.33E+8 8.66E+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 1.98E+2 5.71E+S 2.79E+l 58

Salem ODCM Rev . 6 03/28/90

..

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 6.89E+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 1.11E+7 2.33E+7 Fe-59 2.25E+8 3.93E+8 L 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 3.49E+10 2.16E+9 1.07E+8 1.21E+9 Zn-65 5.55E+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 1.18E+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 Ag-llOm 3.86E+8 2.82E+8 4.03E+8 1.46E+10 1.86E+8 Sb-124 - 2.09E+8 3.08E+6 5.56E+5 1. 31E+8 6.46E+8 6.49E+7 Sb-125 1.49E+8 1.45E+6 1.87E+5 9.38E+7 l.99E+8 3.07E+7 Te-125m 1.51E+8 5.04E+7 5.07E+7 7.18E+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 6.80E+l0 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 5.15E+10 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 1. 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. 6 03/28/90 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

. Nucl.ide 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+5 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 1. 04E+l0 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-89 9.96E+9 1.60E+9 2.86E+8 Sr-90 6.05E+ll 1. 75E+10 1. 48E+ll Y-~1 5.13E+6 2.82E+9 1. 37E+5 Zr-95 l.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-t8 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 l.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 l.39E+5 l.05E+8 l.50E+9 3.25E+7 Te-125m 9.66E+7 3.50E+7 2.90E+7 3.93E+8 3.86E+8 1. 29E+7 Te-127m 3.49E+8 1.25E+8 8.92E+7 1. 42E+9 1._17E.+9 4.26E+7 Te-129m. 2.55E+8 9.50E+7 8.75E+7 1.06E+9 1.28E+9 4.03E+7 I-131 8.09E+7 1.16E+8 3.79E+10 1.98E+8 3.05E+7 6.63E+7 Cs-134 4.66E+9 i:-i1E+lO 3.59E+9 1.19E+9 1. 94E+8 9.07E+9 Cs-136 4.20E+7 1.66E+8 9.24E+7 1.27E+7 1. 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 l.29E+8 1.62E+5 5.49E+4 9.25E+4 2.65E+8 8.43E+6 Ce-141 1. 96E+5 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. 6 03/28/90

  • 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 1. 61E+9 9.96E+7 1.35E+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. 35E+8 9.27E+8 8.97E+7 Fe-55 3.25E+8 2.31E+8 1.46E+8 9.98E+7 5. 3_8E+7 Fe-59 l.81E+8 4.22E+8 1.33E+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 . 1. 01E+8 Co-60 2.49E+8 3.24E+9 5.60E+8 Ni-63 1.61E+10 1.13E+9 1.81E+8 5. 45E+a*

zn-65 4.24E+B 1.47E+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+10 1. 80E+9 4.33E+8 Sr-90 7.51E+ll 2.llE+lO 1. 85E+ll Y-91 7.87E+6 3.23E+9 2.11E+5 Zr-95 1. 74E+6 5.49E+5 8.07E+5 1.27E+9 3.78E+5 Nb-95 1. 92E+5 1. 06E+5- 1. 03E+5 4.55E+8 5.86E+4 Ru-103 6.87E+6 2.42E+7 5.74E+8 2.94E+6 Ru-106 3.09E+8 5.97E+8 1.48E+10 3.90E+7 Ag-llOm 1.52E+7 1.44E+7 2.74E+7 4.04E+9 8.74E+6 Sb-124 1.55E+8 2.85E+6 3.51E+5 1. 35E+8 3.11E+9 6.03E+7 Sb-125 2.14E+8 2.34E+6 2.04E+5 1. 88E+8 1.66E+9 5.00E+7 Te-125m 1.48E+8 5.34E+7 4.14E+7 4.37E+8 1.98E+7 Te-127m 5.51E+8 l.96E+8 1. 31E+8 2.24E+9 1. 37E+9 6.56E+7 Te-129m 3.67E+8 l.36E+8 1.18E+8 1.54E+9 1.38E+9 5.81E+7 I-131 7.70E+7 1.08E+8 3.14~+10 1.85E+8 2.13E+7 5.79E+7 cs-134 7.09E+9 1. 67E+10 5.30E+9 2.02E+9 2.08E+8 7.74E+9 cs-136 4.29E+7 1.69E+8 9.19E+7 1. 45E+7 1.36E+7 l.13E+8 cs-137 1. OlE+lO l.35E+10 4.59E+9 1. 78E+9 1.92E+8 4.69E+9 Ba-140 l.38E+8 1. 6-9E+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 5.27E+7 2.18E+7 1. 30E+7 1.33E+10 2.83E+6 Pr-143 7.12E+4 2.84E+4 1. 65E+4 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. 6 03/28/90

  • 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. OlE+.5

  • 7.01E+5 P-32 3.37E+9 l.58E+8 9.30E+7 1. 30E+8 Cr-51 6.54E+4 1. 79E+4 1.19E+5 6.25E+6 1.18E+5 Mn-54 6.61E+8 1.85E+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 4.01E+8 6.49E+8 1. 88E+8 6.76E+8 3.23E+8 Co-57 2.99E+7 2.45E+8 6.04E+7 Co-58 6.47E+7 3.77E+8 l.98E+8 Co-60 3.78E+8 2.10E+9 l.12E+9 Ni-63 3.95E+10 2.11E+9 1.42E+8 1. 34E+9 Zn-65 8.12E+8 2.16E+9 1. 36E+9 3.80E+8 1. 35E+9 Rb-86 4.52E+8 2.91E+7 2.78E+8 Sr-89 3.59E+10 1.39E+9 1.03E+9 Sr-90 l.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+5 2.95E+8 1.14E+5 Ru-103 1.55E+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 l.05E+8 Te-125m 3.51E+8 9.50E+7 9.84E+7 3.38E+8 4.67E+7 Te-127m 1.32E+9 3.56E+8 3.16E+8 3.77E+9 1.07E+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 1. 44E+8 4.76E+10 2.36E+8 1. 28E+7 8.18E+7 Cs-134 1.60E+10 2.63E+10 8.14E+9 2.92E+9 1. 42E+8 5.54E+9 cs-136 8.06E+7 2.22E+8 1.18E+8 1. 76E+7 7.79E+6 1.43E+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 1. 62E+7 Ce-141 6.53E+5 3.26E+5 1. 43E+5 4.07E+8 4.84E+4 Ce-144 1.27E+8 3.98E+7 2.21E+7 1. 04E+10 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. 6 03/28/90 Table 2-5 (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-51 4.68E+6 Mn-54 1. 34E+9 Fe-55 Fe-59 2.75E+8 Co-58 3.82E+8 Co-60 2.16E+10 Ni-63 Zn-65 7.45E+8 Rb-86 8 .-98E+6 Sr-89 2.16E+4 Sr-90

_Y-91 1.08E+6 Zr-95 2.48E+8 Nb-95 1. 36E+8 Ru-103 l.09E+8 Ru-106 4.21E+8 Ag-llOm 3. 4 7E+9, Te-125m 1. 55E+6 Te-127m 9.17E+4 Te-129m 2.00E+7 I-131 1.72E+7 Cs-134 6.75E+9 Cs-136 1~49E+8 Cs-137 1. 04E+l0 Ba-140 2.05E+7 Ce-141 1.36E+7 Ce-144 6.95E+7 Pr-143 Nd-147 8.40E+6 63

Salem ODCM Rev. 6 3/28/90 APPENDIX A Evaluation of Default MPC Value for Liquid Effluents

  • A-1

Salem ODCM Rev. 6 3/28/90 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 corresponding alarm setpoint is a function of the individual radionuclide distribution and corresponding MPC values.

In order to limit the need for routinely having to reestablish the alarm setpoints as a function of changing radionuclide distributions~ 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. 6 3/28/90 (A. l)

C*1 E --------

MPCi where:

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

= concentration of radionuclide i in the mixture

= the 10 CFR 20, Appendix B, Table II, Column 2 MPC value for radionuclide i (uCi/ml)

Based on the above equation and the radionuclide distribution in the effluents for past years from Salem, an effective MPC value can be determine. Results are presented in Table A-1 and A-2 for Unit 1 and Unit 2, respectively.

Considering the average effective MPC value for the years 1981 through 1989, it is reasonable to select an MPCe value of lE-05 uCi/ml as typical of liquid radwaste discharges. Using this value 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-3

Salem ODCM Rev. 6 3/28/90

  • 4 Table A-1 Calculation of Effective MPC Salem Unit 1 Activity Released (Ci)

Nuclide MPC

  • 1984 1985 1986 1987 1988 1989 CuCi/ml)

--------- --------- --------- --------- -------- --------

Na-24 3E-05 5.6E-03 6.2E-03 9.2E-04 6.9E-04 1.38E-02 4.69E-04 Cr-51 2E-03 5.3E-02 3.6E-02 6.0E-02 N/D 2.38E-02 5.25E-03 Mn-54 1E-04 1.9E-01 8.7E-02 1.9E-01 1.0E-01 1.01E-01 1.12E-01 Fe-59 5E-05 5.8E-03 1.4E-03 2.4E-03 N/D 2.66E-04 1.32E-03 Co-58 9E-05 1.6 6.6E-01 2.22 1.54 1.27E+OO 1.82E+OO Co-60 3E-05 1.2 6.5E-01 3.1E-01 4.2E-01 2.nE-01 1.78E-01 Zr-95 6E-05 1.8E-03 3.2E-03 4.3E-03 8.6E-04 1.23E-02 1.53E-03 Nb-95 1E-04 1. 7E-02 1.3E-03 1.8E-02 2.4E-03 1.53E-02 3.85E-03 Nb-97 9E-04 2.0E-02 7.2E-03 1.5E-03 9.8E-03 2.44E-02 7.94E-05 Tc-99m 3E-03 1.6E-03 N/D N/D 1.1E-04 4.74E-03 4.62E-04 Sr-89

  • 3E-06 4.2E-04 1. 7E-03 3.5E-07 1.6E-02 1.25E-02 9.37E-04 Sr-90 3E-07 2.2E-05 1. 7E-04 3.1E-08 7.7E-04 2.40E-03 3.75E-04 Mo-99 4E-05 1.9E-03 1.0E-04 N/D 1.0E-04 1.57E-03 N/D Ag-110m 3E-05 N/D N/D N/D 2.8E-03 4.96E-03 2.70E-03 Sn-113 8E-05 9.4E-04 N/D 3;5E-04 N/D N/D N/D Sb-124 2E-05 1.7E-02 5.7E-03 8.4E-02 2.4E-02 6.32E-02 1.36E-02 Sb-125 1E-04 4.9E-03 N/D 3.6E-02 3.3E-02 9.35E-02 6.53E-02 1-131 3E-07 4.5E-02 7.9E-02 1.2E-01 1.8E-01 5.54E-02 3.04E-02 1-133 1E-06 1.9E-02 1.4E-03 1.9E-02 2.80E-02 6.88E-03 1-135 4E-06 1.2E-03 N/D N/D 2.0E-03 1.68E-02 1.94E-04 Cs-134 9E-06 5 .1E-02 1.6E-01 3.4E-01 3.1E-03 1.31E-01 1.16E-01 Cs-137 2E-05 5.8E-02 2.1E-01 3.6E-01 3.0E-01 1.34E-01 1.28E-01 Ba-140 2E-05 2.1E-03 N/D N/D N/D 2.79E-04 N/D La-140 2E-05 1.6E-02 1.1E-04 3.5E-04 N/D 3.89E-04 2.66E-04 Total Ci 3.32 1.93 3.75 3.26 2.29E+OO 2.49E+OO c.

-1 2.46E+05 3.42E+05 4.99E+05 7.31E+05 2.80E+05 1.58E+05 MPCi MPCe (UCi/ml) 1.35E-05 5.63E-06 7.51E-06 4.46E-06 8.18E-06 1.58E-05

    • N/D - not detected A-4

~-

Salem ODCM Rev. 6 3/28/90 Table A-2 Calculation of Effective MPC Salem Unit 2 Activity Released CCi)

Nuclide MPC

  • 1984 1985 1986 1987 1988 1989 (uCi/ml)

--------- --*------ --------- --------- -------- --------

Na-24 3E-05 4.4E-03 3.5E-03 3.6E-03 7.3E-05 1.04E-02 8.08E-04 Cr-51 2E-03 3.6E-02 3.5E-02 9.5E-02 3.0E-03 3.17E-03 1.57E-02 Mn-54 1E-04 1.6E-01 1.1E-01 2.2E-01 1.2E-01 1. 74E-01 1.19E-01 Fe-59 5E-05 7.6E-03 1.1E-03 4.0E-03 N/D 2.93E-05 3.00E-03 Co-58 9E-05 1.3 8.4E-01 3.32 1.7 1.32E+OO 2.02E+OO Co-60 3E-05 9.8E-01 6.3E-01 3.8E-01 4.2E-01 2.97E-01 2.08E-01 Zr-95 6E-05 1.2E-03 4.6E-03 1.1E-02 8.4E-04. 3.15E-03 3.39E-03 Nb-95 1E-04 1.4E-02 1.4E-02 2.5E-02 6.6E-03 6.55E-03 7.41E-03 Nb-97 9E-04 2.1E-02 5.7E-03 2.7E-03 N/D 6.92E-03 2.54E-04 Tc-99m 3E-03 1.4E-03 N/D N/D 5.7E-04 3.28E-03 6.64E-04 Sr-89 3E-06 3.2E-04 1.5E-03 -- 4.1E-07 3.0E-03 1.69E-02 9.22E-04 Sr-90 3E-07 4.1E-05 1.0E-04 3.2E-08 2.9E-04 4.11E-03 2.71E-04 Mo-99 4E-05 1.4E-03 N/D N/D 4.4E-04 1.19E-04 N/D

  • Ag-110m 3E-05 N/D N/D N/D N/D 1.04E-02 6.41E-03 Sn-113 8E-05 1.2E-03 N/D 1.1E-03 N/D N/D N/D Sb-124 2E-05 3.0E-02 1.2E-03 1.2E-01 4.6E-02 5.47E-02 1.89E-02 Sb-125 1E-04 3.6E-03 N/D 5.4E-02 5.9E-02 9.22E-02 8.08E-02 I-131 3E-07 4.2E-02 8.4E-02 1.2E-01 2.2E-01 1.35E-01 3.79E-02 1-133 1E-06 2.6E-02 1.2E-02 2.6E-03 1.8E-02 8.83E-02 8.64E-03 I-135 4E-06 4.4E-04 N/D N/D N/D 1.90E-02 5.17E-04 Cs-134 9E-06 2.6E-02 1.8E-01 3.6E-01 3.5E-01 9.53E-02 1.43E-01 Cs-137 2E-05 4.8E-02 2.3E-01 3.7E-01 3.3E-01 1.09E-01 1.55E-01 Ba-140 2E-05 6.6E-03 N/D N/D N/D 1.57E-03 N/D La-140 2E-05 3.0E-02 N/D 6.9E-04 N/D 1.03E-03 5 .19E"04 Total Ci 2.74 2.15 5.09 3.85 2.45E+OO 2.83E+OO c.

-1 2.24E+05 3.56E+05 5.20E+05 8.59E+05 6.09E+05 1.93E+05 MPCi MPCe (uCi/ml) 1.22E-05 6.04E-06 9.79E-06 4.49E-06 4.02E-06 1.47E-05

    • N/D - not detected A-5

~-

Salem ODCM Rev. 6 3/28/90

...

  • APPENDIX B Technical Basis for Effective Dose Factors Liquid Radioactive Effluent
  • B-1

Salem ODCM Rev. 6 3/28/90 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. 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 J.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 radionuclides) 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

Salem ODCM Rev. 6 3/28/90 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 dose 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 e911ations may be used:

Total Body 1.67E-02

  • VOL Dtb = A Fe-59,TB * (B. 1)
  • cw
  • where:

Dtb = dose to the total body (mrem)

A Fe-59,TB = 7.27E+04, total body ingestion dose conversion factor for Fe-59 (mrem/hr per uCi/ml)

VOL = volume of liquid effluent released (gal)

C*l. = 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:

1.21E+03

  • VOL Dtb = cw (B.2)
    • -
  • B-3

Salem ODCM Rev. 6 3/28/90

  • Maximum organ l.67E-02
  • VOL
  • A Nb-95,GI-LLI
  • E C*i (B.3) cw 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 = --------------

cw

  • (B.4)

Tritium is not included in the limited analysis dose assessment for liquid releases, because the potential dose resulting from 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

Table B-1 Adult Dose contributions Fish and Invertebrate Pathways Unit 1 1989 1988 1987 Radio- RELEASE TBOOY GI-LLI LIVER RELEASE TBOOY GI-LLI LIVER RELEASE TBODY 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 0.03 0.06 0.11 1.01E-01 0.01 0.03 0.03 1.05E-01 0.01 0.05 0.-04 FE-55 3.98E-02 0.05 0.02 0.19 5.44E-01 0.43 0.17 0.76 2.35E-01 0.16 0.11 0.43 FE-59 1.32E-03 0.02 0.02 0.03 2.66E-04 * *

  • N/D * *
  • C0-58 1.82E+OO 0.39 0.56 0.16 1.27E+OO 0.17 0.24 0.03 1.54E+OO 0.17 0.42 0.05 C0-60 1.78E-01 0.11 0.15 0.04 2.m-01 0.10 0.14 0.02 4.21E-01 0.13 0.31 0.04 ZN-65 3.62E-04 0.01
  • 0.02 5.49E-04 0.01
  • 0.01 N/D * *
  • NB-95 3.85E-03
  • 0.15
  • 1.53E-02
  • 0.36
  • 2.44E-03
  • 0.08
  • AG-110M 2.70E-03
  • 0.04
  • 4.96E-03
  • 0.05
  • 2.36E-03
  • 0.03
  • CS-134 1.16E-01 0.24
  • 0.25 1.31E-01 0.17
  • 0.08 3.11E-01 0.34
  • 0.21 1.34E-01 0.10
  • 0.06 3.01E-01 0.19
  • 0. 19 Total 2.40E+OO 2.48E+OO 2.92E+OO Table B-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1989 1988 1987 ISOTOPE RELEASE TBOOY GI-LLI LIVER RELEASE TBOOY GI *LLI LIVER RELEASE TBODY GI-LL! 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 0.09 1. 74E-01 0.03 0.07 . 0.06 1.20E-01 0.01 0.04 0.02 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 0.72 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 0.14 1.32E+OO 0.19 0.29 0.04 1. 71E+OO 0.12 0.31 0.02 C0-60 2.0BE-01 0.11 0.13 0.04 2.97E-01 0.12 0.18 0.02 4.23E*01 0.09 0.21 0.02 ZN-65 1.41E-04 * * . 0.01 N/D * *
  • N/D * *
  • NB-95 7.41E-03
  • 0.22
  • 6.55E-03
  • 0.18
  • 7.92E-03
  • 0.18 -...-* .

AG-110M 6.41E-03

  • 0.07
  • 1.04E-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 0.14 * - 0.09 Total 2.71E+OO 2.48E+OO 3.82E+OO
  • less than 0.01

/D = not detected

... Salem ODCM Rev. 6 3/28/90

  • APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent
    • C-1

Salem ODCM Rev. 6 3/28/90

..

  • 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 (Keff 1 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:

Kef f = the effective total body dose factor due to gamma emissions from all noble gases released K*1 = the total body dose factor due to gamma emissions from each noble gas radionuclide i released f 1* = the fractional abundance of noble gas radionuclide i relative to the total noble gas activity C-2

Salem ODCM Rev. 6 3/28/90

  • (C.2) where:

(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 (C. 3) where:

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

  • where:

(C.4)

= the effective air dose factor due to beta emissions

.from all noble gases released

= the air dose factor due to beta emissions from 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. 6 3/28/90 2,800 Ci for 1985, 2,700 Ci for 1986, 1700 Ci for 1988, and 1500 Ci for 1989. Therefore, in order to provide a reasonable basis for r

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 conservatism 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:

3.17E-08 Dg =--* --------

o.so

  • X/Q
  • Mef f
  • E Qi (C. 5) and 3.17E-08

= --------

o.so

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

Salem ODCM Rev. 6 3/28/90

.

  • where:

Dg Db

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

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

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

Mef f = 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 constants, the dose calculational equations simplify to:

= 3.5E-05 X/Q (C.7)

Dg *

  • E Qi and

= 7.0E-05 X/Q (C.8)

  • E Qi 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-5

Salem ODCM Rev. 6 3/28/90 Table C-1 Effective Dose Factors Noble Gases - Total Body and Skin Total Body Effective Skin Effective Radionuclide f.*

1 Dose Factor Dose Factor Ke ff (l+ 1.1 M>eff (mrem/yr per UCi/m3 ) (mrem/yr per uCitm3 )

--------------------

Kr-85 0.01 1.4E+01 Kr-88 0.01 1.5E+02 1.9E+02 Xe-133m 0.01 2.5E+OO 1.4E+01 Xe-133 0.95 3.0E+02 6.6E+02 Xe-135 0.02 3.6E+01 7.9E+01 Total 4.8E+02 9.6E+02 Noble Gases - Air Gamma Air Effective Beta Air Effective Radionuclide f.* Dose Factor Dose Factor 1

"ef f Neff (mrad/yr per uCitm3 ) (mrad/yr per uCitm3 )

-------------------- ---------------------

Kr-85 0.01 2.0E+01 Kr-88 0.01 1.5E+02 2.9E+01 Xe-133m 0.01 3.3E+OO 1.5E+01 Xe-133 0.95 3.4E+02 1.0E+03 Xe-135 0.02 3.8E+01 4.9E+01 Total 5.3E+02 1.1E+03

  • Based on Noble gas distribution from ANSI N237-1976/ANSI-18.1, "Source Term Specifications."

C-6

Salem ODCM Rev. 6 3/28/90

  • APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1

Salem ODCM Rev. 6 3/28/90 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 ~etermining 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~resented 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. Multiplication of the total release (i.e. cumulative activity for all radlonuclides) by this dose conversion factor provides for a dose calculation method that is simplified while also being conservative.

D-2

Salem ODCM Rev. 6 3/28/90 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:

Dmax = 3.17E-8-* W

  • RI-131
  • E Qi where:

Dmax = maximum organ dose (mrem) w = atmospheric dispersion parameters to the controlling location(s) as identified in Table 3.2-4.

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

D/Q = atmosperic deposition for vegetation, milk nad 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

= 1.05El2 (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 fo the overall negligible contribution of these pathways to

  • D-3

Salem ODCM Rev. 6 3/28/90 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 maximize 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 exposure pathways and the maximum organ soe 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 ODCM Rev. 6 3/28/90

..

  • 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 }2y Pathway Pathway Grass-Cow-Milk 0.92 Ground Plane 0.08 Inhalation
  • D-5

Salem ODCM Rev. 6 3/28/90 Appendix E Radiological Environmental Monitoring Program Sample Type, Location and Analysis

  • E-1

. Salem ODCM Rev. 6 3/28/90 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 = Surface 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 = 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 off-site c = 2-3 miles off-site H = > 20 miles off-site 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 J

Salem ODCM Rev. 6 3/28/90.

  • 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 0.14 mi. SSW of vent; site shoreline IDM llSl 0.09 mi. SW of vent; site shoreline IDM llAl 0.2 mi. W of vent; outfall area ECH, ESF, ESS, SWA 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 402 3.7 mi. ENE of vent; Alloway Creek IDM Neck Road 501 3.5 mi. E of vent; local farm AIO, APT, IDM, WWA 1001 3.9 mi. SSW of vent; Taylor's Bridge IDM Spur E-3

Salem ODCM Rev. 6 3/28/90

  • STATION CODE TABLE E-1 (Cont'd)

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 f qarm 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 16El 4.1 mi. NNW of vent; Port Penn AIO, APT, IDM lFl 5.8 mi. N of vent; Fort Elfsborg AIO, APT, IDM

_lF2 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. 6 3/28/90 TABLE E-1 (Cont'd)

STATION CODE STATION LOCATION SAMPLE TYPES 2F6 7.3 mi. NNE of vent; Southern Training IDM Center 2F7 5.7 mi. NNE of vent; local farm MLK, VGT 3F2 5 .1 mi. NE of vent; Hancocks Bridge IDM Municipal Building 3F3 a. 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 7F2 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 IDM 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 MLK, VGT 14F2 6.6 mi. WNW of vent; Boyds Corner IDM 14F3 5.4 mi. WNW of vent; local farm FPV 15F3 5.4 mi. NW of vent IDM E-5

Salem ODCM Rev. 6 3/28/90 TABLE E-1 (Cont'd)

STATION CODE 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 lG3 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 Wilmington IDM Airport 3Hl 32 mi. NE of vent; National Park, NJ IDM 3H3 110 mi. NE of vent; Research and AIO, APT, IDM Testing 3H5 25 mi. NE of vent; local farm FPL, FPV E-6

Salem ODCM Rev. 6 3/28/90 SAMPLES COLLECTION AND ANALYSIS Sample 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 thoron 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 isotopic 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. 6 3/28/90

  • SAMPLE COLLECTION AND ANALYSIS (Cont'd)

Sample 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 providing winter monthly H-3 on Surface) icing conditions allow. quarterly surface sample, monthly on ground water sample .

  • E-8

FJGURE E-1 OFFSITE SAMPLING LOCATIONS 16 2

MANNINGTON 15 14 4

5 12 6

ll 7

8 9

~.

.. FIGURE E-2 ONSITE SAMPLING LOCATIONS 1

15 3

GE 14 Cl::'.'.

w

>

f-1 Cl::'.'.

i 5 M UM EXCLUSI AREA BOUNDA C901 METE )

R ER 6 11 7

10 N 9

J

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