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{{#Wiki_filter:* SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Revision 3 07/30/87 * * ---------------------.., ( 8709030520 870828 I PDR ADOCK 05000272 R PDR
{{#Wiki_filter:*
* *
SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Revision 3 07/30/87
* SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL Table of Contents Introduction  
* (
* . . . . . . . . . . .. 1 1.0 Liquid Effluents 2.0 3.0 1.1 Radiation Honitoring Instrumentation and Controls * * * *
I PDR R
* 2 1.2 Liquid Effluent Honitor Setpoint Determination  
8709030520 870828 ADOCK 05000272 PDR
* * * * *
 
SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL                       )
Table of Contents Introduction *                                   . . . . . . . . . . ..     1 1.0   Liquid Effluents 1.1 Radiation Honitoring Instrumentation and Controls * * * *
* 2 1.2 Liquid Effluent Honitor Setpoint Determination * * * * *
* 3 1.2.1 Liquid Effluent Honitors <Radwaste, Steam Generator Blowdown and Service Water> * * * * * * * * * * *
* 3 1.2.1 Liquid Effluent Honitors <Radwaste, Steam Generator Blowdown and Service Water> * * * * * * * * * * *
* 4 1.2.2 Conservative Default Values *************
* 4 1.2.2 Conservative Default Values       ************* 5 1.3 Liquid Effluent Concentration Limits - 10 CFR 20                   5 1.4 Liquid Effluent Dose Calculations - 10 CFR 50 * * *
5 1.3 Liquid Effluent Concentration Limits -10 CFR 20 5 1.4 Liquid Effluent Dose Calculations  
* 7 1.4.1 Hember of the Public Dose - Liquid Effluents
-10 CFR 50 * * *
* 7 1.4.2 Simplified Liquid Effluent Dose Calculation                 8 1.5 Secondary Side Radioactive Liquid Effluents -
* 7 1.4.1 Hember of the Public Dose -Liquid Effluents
Dose Calculations During Primary to Secondary Leakage *
* 7 1.4.2 Simplified Liquid Effluent Dose Calculation 8 1.5 Secondary Side Radioactive Liquid Effluents  
* 10 1.6 Liquid Effluent Dose Projection * * * * * * * * * * * *       *
-Dose Calculations During Primary to Secondary Leakage *
* 12 2.0  Gaseous Effluents 2.1 Radiation Honitoring Instrumentation and Controls *        * *
* 10 1.6 Liquid Effluent Dose Projection  
* 13 2.2 Gaseous Effluent Honitor Setpoint Determination             * *
* * * * * * * * * * * * *
* 15 2.2.1 Containment and Plant Honitor         ****           * *
* 12 Gaseous Effluents 2.1 Radiation Honitoring Instrumentation and Controls
* 15 2.2.2 Conservative Default Values                             *
* 2.2 Gaseous Effluent Honitor Setpoint Determination 2.2.1 Containment and Plant Honitor **** 2.2.2 Conservative Default Values 2.3 Gaseous Effluent Instantaneous  
* 16 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations - 10 CFR 20 * * * * * * *
* *
* 18 2.3.1 Site Boundary Dose Rate - Noble Gases             ****     18 2.3.2 Site Boundary Dose Rate - Radioiodine and Particulates
* 13 * *
* 19 2.4 Noble Gas Effluent Dose Calculations - 10 CFR 50         ****     21 2.4.1 UNRESTRICTED AREA Dose - Noble Gases * * * * * * * * *
* 15 * *
* 21 2.4.2 Simplified Dose Calculation for Noble Gases                 21 2.5 Radioiodine and Particulate Dose Calculations - 10 CFR 50
* 15 *
* 23 2.5.1 UNRESTRICTED AREA Dose - Radioiodine and Particulates
* 16 Dose Rate Calculations  
* 23 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates . * * * * * * * * * * * * * * * *
-10 CFR 20 * * * * * * *
* 24 2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations * * * * * * * * * * * *
* 18 2.3.1 Site Boundary Dose Rate -Noble Gases **** 18 2.3.2 Site Boundary Dose Rate -Radioiodine and Particulates
* 25 2.7 Gaseous Effluent Dose Projection * * * * * * * * * *          *
* 19 2.4 Noble Gas Effluent Dose Calculations  
* 28 3.0  Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY                   29 3.2 Doses to HEHBERS OF THE PUBLIC - 40 CFR 190 *
-10 CFR 50 **** 21 2.4.1 UNRESTRICTED AREA Dose -Noble Gases **********
* 30 3.2.1 Effluent Dose Calculations **                               30 3.2.2 Direct Exposure Determination                       * *
21 2.4.2 Simplified Dose Calculation for Noble Gases 21 2.5 Radioiodine and Particulate Dose Calculations  
* 31 4.0 Radiological Environ*ental Monitoring Program             * * * * *
-10 CFR 50
* 32 4.1 Sampling Progra* * * * * * * * * *
* 23 2.5.1 UNRESTRICTED AREA Dose -Radioiodine and Particulates
* 32 4.2 Interlaboratory Comparison Program                 . * . . . . . 33
* 23 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates  
 
.****************
                                                              ---Salem ODCH Rev. 3 07/30/87 Table of Contents - Continued Tables .
2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations  
1-1   Para*eters for Liquid Alarm Setpoint Determination - U~it 1 **       37 1-2    Para*eters for Liquid Alarm Setpoint Determination - Unit 2 **       38 1-3    Site Related Ingestion Dose Commitment Factors. Aio * * * * *
************
* 39 1-4    Bioaccumulation Factors <BFi> ~ * * * * * * * * * * * * * * *
2.7 Gaseous Effluent Dose Projection  
* 41 2-1    Dose Factors for Nob 1e Gases * * * * * * * * * * * * * * * *
**********
* 44 2-2    Para*eters for Gaseous Alarm Setpoint Determinations - Unit 1
Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY 3.2 Doses to HEHBERS OF THE PUBLIC -40 CFR 190
* 45 2-3    Parameters for Gaseous Alarm Setpoint Determinations - Unit 2
* 3.2.1 Effluent Dose Calculations  
* 46 2-4    Controlling Locations. Pathways and Atmospheric Dispersion for Dose Calculations * * * * * * * * * * * * * * * * *
** 3.2.2 Direct Exposure Determination
* 49 2-5    Path1111ay Dose Parameters - Atmospheric Re.leases * * * *       **  so A-1    Calculation of Effective HPC - Unit 1 * * * * *
* 24
* A-4 A-2    Calculation of Effective HPC - Unit 2 * * * * * * . * * *
* 25 *
* A-5 B-1    Adult Dose Contribut.ions Fish and Drinking Water Pathways Unit 1 * * * * . * * * * * *         * * * * * * * * * * * *
* 28 29
* B-5 B~2    Adult Dose Contributions Fish and Drinking Water Pathways Unit 2 * * * * * * * * * * * * *         * * * * . * . * *
* 30 30 * *
* B-5 C-5   Effective Dose Factors               * * * * * * * * * * *
* 31 4.0 Radiological Environ*ental Monitoring Program 4.1 Sampling Progra* *********  
* C-5 Appendic ies Appendix A - Evaluation of Conservative. Default HPC Value for Liquid Effluents * * * * *
* * * * *
* A-1 Appendix B Technical Basis for Effective Dose Factors -
* 32
Liquid Radioactive Effluents * * * * * * * * * * *
* 32 4.2 Interlaboratory Comparison Program . * . . . . .
* B-1 Appendix C - Technical Bases for Effective Dose Factors ~
33 ) 
Gaseous Radioactive Effluents                       ** C~1 Appendix D - Radiological Environmental Monitoring Program -
* * * ---Salem ODCH Rev. 3 07/30/87 Table of Contents -Continued Tables . 1-1 1-2 1-3 1-4 2-1 2-2 2-3 2-4 2-5 A-1 A-2 B-1 Para*eters for Liquid Alarm Setpoint Determination  
Sample Type. Location and Analysis * * * *       * *
-
* D-1
1 ** 37 Para*eters for Liquid Alarm Setpoint Determination  
 
-Unit 2 ** 38 Site Related Ingestion Dose Commitment Factors. Aio ****** 39 Bioaccumulation Factors <BFi> ****************
Salem ODCH Rev. 3 07130187 SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL
41 Dose Factors for Nob 1 e Gases * * * * * * * * * * * * * * * *
* The   Salem Offsite Dose Calculation Hanual COOCH> describes the methodology parameters   used   in!   1)   the calculation of radioactive   liquid   and 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 501   Appendix   I and 40 CFR 190.
* 44 Para*eters for Gaseous Alarm Setpoint Determinations  
Hore 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 *
-Unit 1
* The made ODCH training to will be maintained at the station for use as a document of accepted methodologies and calculations.
* 45 Parameters for Gaseous Alarm Setpoint Determinations  
the ODCH   calculation methodologies and reference parameters Chang~s as   is guide  and will be 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.
-Unit 2
NOTE!     As used throughout this document. excluding acronyms. words appearing all capitalized denote the application of definitions as used in the Salem Technical Specifications *
* 46 Controlling Locations.
* 1
Pathways and Atmospheric Dispersion for Dose Calculations  
 
******************
Salem ODCH Rev. 3 07130187
Path1111ay Dose Parameters  
* The liquid controlling effluent monitoring   instrumentation and controls and monitoring normal radioactive material relea*ses at in Salem  for accordance with the Salem Radiological Effluent Technical Specifications are summarized as follows:
-Atmospheric Re.leases  
1>   81Dcm_ian~_8Y12maii~-I~cminaii2n1 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.
* Calculation of Effective HPC -Unit 1 ***** Calculation of Effective HPC -Unit 2 ******.***
1-R19 A181C1and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R19 A181C and D provide this function for Unit 2.
Adult Dose Contribut.ions Fish and Drinking Water Pathways Unit 1 * * * * . * * * * * * * * * * * * * * * * *
: 2) 8lDcm_i2nl~1 - The alarm functions ror the Service Water System are provided by the radiation monitors on the Containment Fan Cooler discharges C1-R 13 A181C1D and E for Unit 1 and 2-R 13 A181and C for Unit 2>.
* 49 ** so
Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored Cwith an alarm function>
* A-4
* A-5
* B-5 Adult Dose Contributions Fish and Drinking Water Pathways Unit 2 * * * * * * * * * * * * * * * * * . * . * *
* B-5 C-5 Effective Dose Factors * * * * * * * * * * *
* C-5 Appendic ies Appendix A -Evaluation of Conservative.
Default HPC Value for Liquid Effluents  
* * * * *
* A-1 Appendix B Technical Basis for Effective Dose Factors -Liquid Radioactive Effluents  
************
B-1 Appendix C -Technical Bases for Effective Dose Factors Gaseous Radioactive Effluents Appendix D -Radiological Environmental Monitoring Program -Sample Type. Location and Analysis **** ** * *
* D-1
* *
* SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL Salem ODCH Rev. 3 07130187 The Salem Offsite Dose Calculation Hanual COOCH> 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 501 Appendix I and 40 CFR 190. Hore 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 ODCH will be maintained at the station for use as a reference guide and training document of accepted methodologies and calculations.
will be made to the ODCH 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 ODCH Rev. 3 07130187 The liquid effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material relea*ses in accordance with the Salem Radiological Effluent Technical Specifications are summarized as follows: 1> 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 A181C1and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R19 A181C and D provide this function for Unit 2. 2)  
-The alarm functions ror the Service Water System are provided by the radiation monitors on the Containment Fan Cooler discharges C1-R 13 A181C1D and E for Unit 1 and 2-R 13 A181and C for Unit 2>. Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored Cwith an alarm function>
by R37 prior to release to the environ*ent.
by R37 prior to release to the environ*ent.
Liquid radioactive waste flow diagrams with the applicable, associated radiation  
Liquid radioactive waste flow diagrams with the applicable, associated radiation
*onitoring instrumentation and controls are presented ae Figures 1-1 and 1-2 for Units 1 and 2, respectively
  *onitoring instrumentation and controls are presented ae Figures 1-1 and 1-2 for Units 1 and 2, respectively *
* 2
* 2
* *
 
* Salem ODCH Rev. 3 07/30/87 Per the require*ents of Technical Specification 3.3.3.81 alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of Specification 3.11.1.1 are met (i.e ** the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 21 for radionuclides and 2.0E-04 uCi/al for dissolved or entrained noble gases>. The following equation*  
Salem ODCH Rev. 3 07/30/87 Per the require*ents of Technical Specification 3.3.3.81       alarm setpoints   shall be established for the liquid effluent monitoring instrumentation to ensure that the   release concentration limits of Specification 3.11.1.1 are met     (i.e **   the concentration     of   radioactive material   released in   liquid effluents     to UNRESTRICTED     AREAS shall be limited to the concentrations specified in 10       CFR 20,   Appendix B,   Table II, Column 21 for radionuclides and 2.0E-04 uCi/al for dissolved or entrained noble gases>.     The following equation* *ust be aatisfied to meet the liquid effluent restrictions:
*ust be aatisfied to meet the liquid effluent restrictions:
C <F+f)                                 (1.1) c { -------
where: c = c = f = C <F+f) (1.1) c { -------' the effluent concentration limit of Technical Specification (3.11.1.1>
where:
implementing the 10 CFR 20 HPC for the site. in uCi/ml the setpoint1 in uCi/ml1 of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and subsequent release: the setpoint.
c  =  the   effluent concentration limit of       Technical   Specification (3.11.1.1> implementing the 10 CFR 20 HPC for the site. in uCi/ml c  =  the setpoint1 in uCi/ml1 of the radioactivity monitor measuring the radioactivity concentration in the effluent line prior to dilution and   subsequent release: the setpoint. represents a value which.
represents a value which. if exceeded.
if exceeded. would result in concentrations exceeding the li*its of 10 CFR 20 in the UNRESTRICTED AREA f  =  the flow rate at the radiation monitor location. in volume per unit ti*e* but in the sa*e units as F, below F = the dilution water flow rate as measured prior to the release point.
would result in concentrations exceeding the li*its of 10 CFR 20 in the UNRESTRICTED AREA the flow rate at the radiation monitor location.
in volu*e per unit ti*e
in volume per unit ti*e* but in the sa*e units as F, below F = the dilution water flow rate as measured prior to the release point. in volu*e per unit ti*e [Note that if no dilution is provided, c{ C. Also. note that when <F> is large co*pared to (f), then <F + f) = F.l
[Note that if no dilution is provided, c{ C.     Also. note that when <F> is large co*pared to (f), then <F + f)   = F.l
* Adapted fro* NUREG-0133 3 
* Adapted fro* NUREG-0133
* *
* 3
* Salem Rev. 3 01130187 -1. 2.1 f
: 1. 2.1     ~i~~id_Ef f l~~ni_M2nii2cm_iBad~aai~L-Si~am_~~n~cai2c_e12~d2wnL-~h~mi,al The setpoints for the liquid effluent monitors
The setpoints for the liquid effluent monitors at the Sale* Nuclear Generating Station are determined by the following equations:
* at   the equations:
where: SP = HPCe = HPCe
Sale* Nuclear Generating Station SP ~
are HPCe
* SEN
* SEN
* CW SP ---------------
* CW determined
+ bkg RR ( 1.2) Ni th: [. Ci HPCe = ------------( 1.3) -( c* J L alar* setpoint corresponding to the maximum allowable release an effective HPC value for the mixture of radionuclides in effluent strea* CuCi/*1> rate (cpm) the Ci = the concentration of radionuclide i in the liquid effluent CuCi/*l>*
                                                          + bkg by  the  following
HPCi = SEN = cw = RR = bkg = *NOTE The concentration  
( 1.2)
*ix must include the *ost recent co*posite of alpha e**itters1 Sr-89, Sr-901 Fe-55, and H-3 per Technical Specification 3.11.1.1
RR Ni th:
* the HPC value corresponding to radionuclide i fro* 10 CFR 201 Appendix 81 Table 11, Colu*n 2 <uCi/*1> the sensitivity value to which the monitor is calibrated (cp* per uCi/*l> the circulating water flow rate (dilution water flow> at the tiae of release (gal/*in) the liquid effluent release rate (gal/*in) the background of the *onitor Ccp*) 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 strea* flow <RR> may be necessary during these periods to meet the discharge criteria.
[. Ci HPCe   =------------                                 ( 1.3)
However. in order to maximize the available plant discharge dilution and thereby minimize the potential offsite doses, releases fro* either Unit-1 or Unit-2 may be routed to either the Unit-1 or Unit-2 Circulating Water System discharge.
J where:
This routing is possible via interconnections between the Service Water Systems <see Figures 1 and 2>
L-     (
* 4
                                                  -~;~:--;
* *
c*
* Salem ODCH Rev. 3 07/30/87 Procedural restrictions prevent simultaneous releases fro* either a single unit or both units into a single Circulating Water Syste* discharge
SP    =  alar* setpoint corresponding to the maximum allowable release rate (cpm)
* 1.2.2 Conservative alarm aetpoints may be deter*ined through the use of default para*eters.
HPCe  =  an effective HPC value for the mixture of radionuclides in the effluent strea* CuCi/*1>
Tables 1-1 and 1-2 su*marize all current default valuea in use for Salem Unit-1 and Unit-21 respectively.
Ci     = the concentration of radionuclide i in the liquid effluent CuCi/*l>*
                  *NOTE       The concentration *ix must include the
                              *ost recent co*posite of alpha e**itters1 Sr-89, Sr-901 Fe-55, and H-3 per Technical Specification 3.11.1.1 *
* HPCi SEN cw =
              =
              =
the HPC value corresponding to radionuclide i fro* 10 CFR 201 Appendix 81 Table 11, Colu*n 2 <uCi/*1>
the     sensitivity value to which the monitor is (cp* per uCi/*l>
the circulating water flow rate (dilution water flow> at the tiae of release (gal/*in) calibrated RR    =  the liquid effluent release rate (gal/*in) bkg  =  the background of the *onitor Ccp*)
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 strea* 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 fro*   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> *
* 4
 
Salem ODCH Rev. 3 07/30/87 Procedural   restrictions prevent simultaneous releases fro* either     a single unit or both units into a single Circulating Water Syste* discharge
* 1.2.2                                     Conservative   alarm   aetpoints   may be deter*ined through the use of default para*eters. Tables 1-1 and 1-2 su*marize all   current default valuea in use for Salem Unit-1   and   Unit-21   respectively.
They are based upon the following:
They are based upon the following:
a> b) c> d) substitution of the effective HPC value with a default value of lE-05 uCi/*1 for radwaste releases (refer to Appendix A for justification>:
a> substitution of the effective HPC value with a default value of lE-05 uCi/*1 for radwaste releases (refer to Appendix A for justification>:
for additional substitution of the 1-131 HPC value 3E-07 uCi/ml for the R19 Stea* Generator blowdown monitors*
b)  for additional ~onservatism** substitution of the 1-131 HPC value of 3E-07 uCi/ml for the R19 Stea* Generator blowdown monitors* R13 Service Water monitor and R37 Che*ical Waste Basin *onitor:
Service Water monitor and R37 Che*ical Waste Basin *onitor: substitutions of the operational circulating water flow with the lowest flow. in gal/*in: and, substitutiona of the effluent release rate with the highest allowed rate. in gal/*in. of R13 With pre-established alarm setpoints.
c>  substitutions of the operational circulating water flow with the lowest flow. in gal/*in: and, d)  substitutiona of the effluent release rate with the highest allowed rate. in gal/*in.
it is possible to control the radwaste release rate <RR> to ensure the inequality of equation <1.2> is *aintained under changing values for HPCe and for differing Circulating Water System dilutions.
With   pre-established   alarm setpoints. it is possible to control the     radwaste release rate <RR> to ensure the inequality of equation <1.2> is *aintained under changing values for HPCe and for differing Circulating Water System dilutions.
Technical Specification 3.11.1.1 li*its 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 201 Appendix B. Table 111 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
Technical   Specification   3.11.1.1 li*its 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 201     Appendix B. Table 111 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
* Use of the effective HPC value as derived in Appendix conservative for the R19 Steam Generator blowdown monitors Waste Basin *onitors where 1-131 transfer during primary to may potentially be *ore controlling
* Use of the effective HPC value as derived in Appendix A may be non-conservative for the R19 Steam Generator blowdown monitors and R37 Chem i cal Waste Basin *onitors where 1-131 transfer during primary to secondary leakage may potentially be *ore controlling *
* 5 A may and R37 be Chem i cal secondary leakage 
* 5
* *
 
* Salem ODCM Rev. 3 07130187 radiation monitor alarm setpoints are established as addressed above to ensure that these concentration limits are not exceeded.
Salem ODCM Rev. 3 07130187 radiation   monitor alarm setpoints are established as addressed above to   ensure
However. in the event any liquid release results in an alarm setpoint being exceeded, an evaluation of co*Pliance with the concentration limits of Techn.ical Specification 3.11.1.1 may be performed using the following equation: 1 ( 1.4) where: Ci = actual concentration of radionuclide as measured in the undiluted liquid effluent (uCi/*l> HPCi = the HPC value corresponding to radionuclide from 10 CFR 20, Appendix e, Table lls Colu*n 2 <uCi/ml> = 2E-04 uCi/*l for dissolved or entrained noble gases RR = the actual liquid effluent release rate (gal/*in>
* that liquid these concentration limits are not exceeded.
CW = the actual circulating water flow rate (dilution water flow) at the ti*e of the release <gal/min) 6
release However.
* *
results in an alarm setpoint being exceeded, in the event an evaluation co*Pliance with the concentration limits of Techn.ical Specification 3.11.1.1 may any of be performed using the following equation:
* Salem ODCH Rev. 3 07/30/87 1.4.1 Technical Specification
                                                          ~  1                   ( 1.4) where:
-3.11.1.2 limits the dose or dose commitment to HEHBERS OF THE PUBLIC fro* radioactive materials in liquid effluents from each unit of the Sale11 Nuclear Generating Station to: during any calendar quarter: 1. S 11re11 to total body per unit 5.0 mrem to any organ per unit during any calendar year: 3.0 11re11 to total body per unit 10.0 mrem to any organ per unit. Per the surveillance require11ents of Technical Specification 4.11.1.2.
Ci     = actual   concentration of radionuclide       as measured in the undiluted liquid effluent (uCi/*l>
the following calculation 11ethode 11ay be used for deter11ining the dose or dose co1111it11ent due to the liquid radioactive effluents fro* Sale11. where: Do = Aio = Ci = VOL = cw = 1.67E-02
HPCi   = the HPC value corresponding to radionuclide     from 10 CFR 20, Appendix e, Table lls Colu*n 2 <uCi/ml>
* VOL Do = --------------
              = 2E-04 uCi/*l for dissolved or entrained noble gases RR     = the actual liquid effluent release rate (gal/*in>
CW * [ .CCi
CW     = the actual circulating water flow rate (dilution water flow) at the ti*e of the release <gal/min)
* Aio) ( 1.5) dose or dose com11it11ent to organ O* including total body (mrem> site-related ingestion dose co*11itment factor to.the total body or any organ o for rad i onuc l i.de i ( 11re11/hr per uC i /111 l > average concentration of radionuclide i* in undiluted liquid effluent representative of the volu11e VOL CuCi/111) voluae of liquid effluent released (gal> average circulating water discharge rate during release period Cga1/11in) 1.67E-02 = conversion factor Chr/11in)
* 6
The site-related ingestion dose/dose commitment factors CA > are presented in io Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:
 
7
Salem ODCH Rev. 3 07/30/87
* *
* 1.4.1 3.11.1.2 radioactive limits     the materials dose or dose commitment to HEHBERS Technical OF   THE in liquid effluents from each unit of the Sale11 Specification PUBLIC    fro*
* Salem Rev. j Aio = 1.14E+05 CCUI
Nuclear Generating Station to:
during any calendar quarter:
            ~ 1. S 11re11 to total body per unit
            ~ 5.0 mrem to any organ per unit during any calendar year:
            ~ 3.0 11re11 to total body per unit
            ~ 10.0 mrem to any organ per unit.
Per   the   surveillance require11ents of     Technical   Specification       4.11.1.2.       the following     calculation     11ethode 11ay be used for deter11ining the dose           or   dose co1111it11ent due to the liquid radioactive effluents fro* Sale11.
1.67E-02
* VOL Do = --------------     *[   .CCi
* Aio)                   ( 1.5)
CW where:
Do      = dose or dose com11it11ent to organ O* including total body (mrem>
Aio      = site-related ingestion dose co*11itment factor to.the total body or any organ o for rad i onuc l i.de i ( 11re11/hr per uC i /111 l >
Ci      = average concentration         of radionuclide i* in undiluted               liquid effluent representative of the volu11e VOL CuCi/111)
VOL      = voluae of liquid effluent released (gal>
cw      = average circulating water discharge rate during release                   period Cga1/11in) 1.67E-02     =   conversion factor Chr/11in)
The   site-related ingestion dose/dose commitment factors CA               > are presented in io Table 1-3 and have been derived in accordance with           of NUREG-0133 by the equation:
* 7
 
Salem UU~M Rev. j   U//jU/~/
* where:
Aio   =
Aio  =  1.14E+05 CCUI
* Bii> + CUF
* Bii> + CUF
* Bfi)l DFi ( 1.6) where: Aio = composite dose parameter for the total body or critical organ o 1.14E+05 UI = Bii = UF = BFi = DFi = of an adult for radionuclide i1 for the fish and invertebrate ingestion pathways <*rem/hr per uCi/*l> = conversion factor CpCi/uCi
* Bfi)l DFi composite dose parameter for the total body or critical organ o of an adult for radionuclide i1 for the fish and invertebrate ingestion pathways <*rem/hr per uCi/*l>
* ml/kg + hr/yr) adult invertebrate consumption CS kg/yr) bioaccu*ulation factor for radionuclide i in invertegrates from Table 1-4 CpCi/kg + pCi/l) adult fish consu11ption  
( 1.6) 1.14E+05    = conversion factor CpCi/uCi
<21 kg/yr) bioaccumulation factor for radionuclide in fish from Table 1-4 CpCi/kg per pCi/1) dose conversion factor for nuclide i for adults in pre-selected organ, 01 from Table E-11 of Regulatory Guide 1.109 <mrem/pCi)
* ml/kg + hr/yr)
The radionuclides included in the periodic dose assessment per the requirements of Technical Specification 3/4.11.1.2 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. In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following si*plified dose calculational equation may be used for demonstrating compliance with the dose limits of Technical Specification 3.11.1.2.  
UI      =  adult invertebrate consumption CS kg/yr)
<Refer to Appendix B for the derivation and justification for this simplified method.> 1.21E+03
Bii    =  bioaccu*ulation factor for radionuclide i in invertegrates from Table 1-4 CpCi/kg + pCi/l)
* VOL Dtb = --------------
UF      =  adult fish consu11ption <21 kg/yr)
* cw ( 1. 7) L_ Ci 8
BFi    =  bioaccumulation factor for radionuclide       in fish from Table 1-4 CpCi/kg per pCi/1)
* where: Ci = VOL = cw = Dtb -D*ax = 1.21E+D3 = 2.52E+D4 = *
DFi    =  dose conversion factor for nuclide i for adults in pre-selected organ, 01 from Table E-11 of Regulatory Guide 1.109 <mrem/pCi)
* Salem ODCH Rev. 3 07/30/87 2.52E+D4
The radionuclides included in the periodic dose assessment per the requirements of Technical Specification 3/4.11.1.2 ~re 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.
* VOL D*ax = --------------
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.
* ( 1.8) cw average concentration of radionuclide i* in undiluted liquid effluent representative of the volu*e VOL <uCi/*l> volu*e of liquid effluent released (gal> average circulating water discharge rate during release period (gal/*in) conservatively evaluated total body dose <*re*> conservatively evaluated  
In lieu of the   individual radionuclide   dose   assessment   as presented in Section   1.4.1,   the following si*plified   dose calculational 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.>
*axi*u* organ dose <*re*> conversion factor (hr/min) and the conservative total body dose conversion factor*<Fe-59, total body --7.27E+04 *rem/hr per uCi/*l > conversion factor <hr/*in) and the conservative 11axi11tu*
1.21E+03
organ dose conversion factor <Nb-95, GI-LLI --1.51E+06 *re*/hr per uC il*l > 9
* VOL Dtb = --------------
* *
cw
* Salem OOCM Rev. 3 07/30/87 1.5
* L_ Ci                    ( 1. 7)
__
* 8
During periods of primary to secondary leakage (i.e ** steam generator tube leaks>. radioactive material will be transmitted from the primary system to the secondary system. The potential exists for the release of radioactive material to the off-site environment  
 
<Delaware River*) via secondary system discharges.
Salem ODCH Rev. 3 07/30/87
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.
* D*ax 2.52E+D4
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 Cheaical Waste Basin with the major source of activity being the Steam Generator blowdown.
* VOL
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). Because the release rate from the secondary system is indirect (e.g., SG blowdown is nor*ally routed to condenser where the condensate clean-up syste* will reaove much of the radioactive  
cw
*aterial>*
                                                *                           ( 1.8) where:
samples should be collected from the final releage point Che*ical Waste Basin) for quantifying the radioactive  
Ci      = average concentration of radionuclide i* in undiluted liquid effluent representative of the volu*e VOL <uCi/*l>
*aterial releases.
VOL      = volu*e of liquid effluent released (gal>
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 10 
cw      = average circulating water discharge rate during release period (gal/*in)
* *
Dtb      - conservatively evaluated total body dose <*re*>
* Salem ODCH Rev. 3 07/30/87 the environment via the Chemical Waste Basin. This approach while not exact, is conservative and ensures timely analysis for regulatory compliance.
D*ax    = conservatively evaluated *axi*u* organ dose <*re*>
Accounting for radioactive material retention of the condensate clean-up system ion exchange resins *BY be needed to more accurately account for actual releases
1.21E+D3 = conversion factor (hr/min) and the conservative total body dose conversion factor*<Fe-59, total body -- 7.27E+04 *rem/hr per uCi/*l >
* 11
2.52E+D4 = conversion factor <hr/*in) and the conservative 11axi11tu* organ dose conversion factor <Nb-95, GI-LLI -- 1.51E+06 *re*/hr per uC il*l >
* *
* 9
* Salem OOCH Rev. 3 07/30/87 Technical Specification 3.11.1.3 requires that the liquid radioactive waste processing syste* be used to reduce the radioactive material levels in the liquid waste prior to release when the quarterly projected doses exceed: 0.375 *re* to the total body1 or 1.25 mre* 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 uaed in conjunction or in lieu of the ion exchange syste* for waate processing require-ments in accordance with Technical Specification 3.11.1.3.
 
Theae processing require*ents are applicable to each unit individually.
Salem OOCM Rev. 3 07/30/87 1.5 S~'2ndac~--Sid~ __ Badi2a,ii~~-Li~~id_E&#xa3;&#xa3;l~~nia_and_Q2m~-~al,~laii2na_Q~cing ecimac~--i2_S~'2ndac~-L~akag~
Exceeding the projected dose requiring processing prior to release for one unit does not in itself dictate processing require*ents for the other unit. Dose projections are *ade at least once per 31 days by the following equations:
* During leaks>.
Dtbp = Otb (91 t d) (1.9) Omaxp = Dmax <91
periods     of primary to secondary leakage     (i.e **   steam   generator radioactive material will be transmitted from the primary system to the secondary system.       The potential exists for the release of radioactive material tube to the   off-site environment <Delaware River*) via secondary system           discharges.
* d> (1.10> where: Dtbp = the total body doae projection for current calendar quarter <*re*> Dtb = the total body dose to date for current calendar quarter as determined by equation <1.S> or <1.7> <*re*> Dmaxp = the maximum organ dose projection for current calendar quarter <*re*> Dmax = the maximum organ dose to date for current calendar quarter as deter*ined by equation (1.5) or (1.8) <mre*> d = the number of days to date for current calendar quarter 91 = the number of days in a calendar quarter 12
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   Cheaical     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 radionuclides.
* Salem ODCM Rev. 3 07/30/87 / The gaseous effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Radiological Technical Specifications are summarized as follows! 1>  
be Based collected on and analyzed for the   principal the identified radioactive material levels gamma    emitting and   the volume of water discharged,       the resulting environmental doses may be calculated based on equation <1.5).
-The vent header gases are collected by the waste*gas holdup system. Gases may 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.
Because   the     release   rate from the secondary   system     is   indirect   (e.g.,   SG blowdown   is   nor*ally routed to condenser where the condensate clean-up           syste*
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>  
will reaove much of the radioactive *aterial>*         samples should be collected from the   final   releage   point (i.e.~  Che*ical Waste   Basin)     for quantifying     the radioactive       *aterial   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
-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 1-R12A and 2-R12A may be used to provide the containment.monitoring and automatic isolation function during purge and pressure/vacuum reliefs.*
* 10
3>  
 
-The Plant Vent for each respective unit receives discharges from the waste gas hold-up systems condenser evacuation system, containment purge and pressure/vacuum reliefs, and the Auxiliary Building ventilation.
Salem ODCH Rev. 3 07/30/87 the environment via the Chemical Waste Basin. This approach while not exact, is conservative and ensures timely analysis for regulatory compliance.
Effluents are monitored by R41C, a flow through gross activity monitor (for noble gas monitoring).
Accounting for radioactive material retention of the condensate clean-up system   ion exchange resins *BY be needed to more accurately account for actual releases *
Additionally, in-line gross activity monitors <1-R16 and 2-R16) provide redundant back-up capabilities to the R41C monitors.
* 11
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  
Salem OOCH Rev. 3 07/30/87
<e.g., venturi rotometer>.
* Technical processing Specification syste*    be 3.11.1.3 requires that the     liquid used to reduce the radioactive material radioactive liquid waste prior to release when the quarterly projected doses exceed:
* The R12A monitors also provide safety function of containment isolation in the event of a fuel handling accident during refueling.
levels    in waste the 0.375 *re* to the total body1 or 1.25 mre* to any organ.
During HOOE 6 in accordance with Technical Specification 3/4.3.3, Table 3.3-6, the R12A alarm/trip setpoint shall be established at twice backgrounds providing early indication and containment isolation accompanying unexpected increases in containment airborne radioactive material levels indicative*
The   applicable   liquid   waste   processing system   for   maintaining   radioactive material releases ALARA is the ion exchange system as delineated in Figure               1-3.
of a fuel degradation.
Alternately,   the waste evaporator as presented in the Salem FSAR has processing capabilities   meeting   the   NRC   ALARA design requirements and may     be   uaed   in conjunction or in lieu of the ion exchange syste* for waate processing             require-ments   in accordance with Technical Specification 3.11.1.3.           Theae   processing require*ents are applicable to each unit individually.           Exceeding the projected dose   requiring   processing   prior to release for one unit does     not   in   itself
The R41C monitor may provide this function if the. R12A monitor is inoperable during HOOE 6
* dictate processing require*ents for the other unit.
* 13
Dose projections are *ade at least once per 31 days by the following equations:
* *
Dtbp   = Otb (91 t d)                               (1.9)
* ODCM Rev. 3 07/30/87 A gaseous radioactive waste flow diagrams with the applicable.
Omaxp   = Dmax <91
associated radiation monitoring instrumentation and controls are presented as Figures 2-1 and 2-2 for Units 1 and 2, respectively
* d>                           (1.10>
* 14
where:
* *
Dtbp   =   the total body doae projection for current calendar quarter
* Salem ODCH Rev. 3 07/30/87 2.2.1 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 calculational method. The measured radionuclide concentrations and release rate are used to calculate the fraction of the allowable release rate, as li*ited by Specification 3.11.2.1, by the equation:
                    <*re*>
Dtb     =   the total body dose to date for current calendar quarter as determined by equation <1.S> or <1.7> <*re*>
Dmaxp   =   the maximum organ dose projection for current calendar quarter
                    <*re*>
Dmax     =   the maximum organ dose to date for current calendar quarter as deter*ined by equation (1.5) or (1.8) <mre*>
d       =   the number of days to date for current calendar quarter 91       =   the number of days in a calendar quarter
* 12
 
Salem ODCM Rev. 3 07/30/87
*                                  /
The   gaseous   effluent   monitoring   instrumentation and controls at Salem for controlling   and monitoring normal radioactive material releases     in accordance with   the   Radiological   Effl~ent    Technical Specifications are summarized as follows!
1>   Waait_Gaa_H2l~YQ-~~&sect;itm - The vent header gases are collected by     the waste*gas holdup system.       Gases may ~e 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> C2niDinmtni_fycgt_Dn~_fctaaYctl~D,YYm-Btlitf - 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>   flDDi-~tni    - The Plant Vent for each respective unit receives discharges from the waste gas hold-up systems 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 2-R16) provide redundant back-up monitor~ng 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>.
* The R12A monitors also provide ~he safety function of containment isolation in the event of a fuel handling accident during refueling.         During HOOE 6 in accordance with Technical Specification 3/4.3.3, Table 3.3-6, the R12A alarm/trip setpoint shall be established at twice backgrounds providing early indication and containment isolation accompanying unexpected increases in containment airborne radioactive material levels indicative* of a             fuel degradation. The R41C monitor may al~o provide this function if the. R12A monitor     is inoperable during HOOE 6 *
* 13
 
Sal~m ODCM Rev. 3 07/30/87 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 *
* 14
 
Salem ODCH Rev. 3 07/30/87 2.2.1                                             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 calculational method.     The measured radionuclide   concentrations and   release   rate are used to calculate the fraction of the     allowable     release rate, as li*ited by Specification 3.11.2.1, by the equation:
FRAC  = C4.72E+02
* X/Q
* VF*  L CCi
* Ki)J + 500              <2.1>
FRAC = C4.72E+02
FRAC = C4.72E+02
* X/Q *VF* [. CCi *<Li+ 1.1 Hi))J + 3000              <2.2>
* where:
FRAC    = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate X/Q    =  annual  average meteorological dispersion to the controlling site boundary location (sec/m3)                                          ,
VF      =  ventilation  system  flow rate for the applicable  release  point  and monitor (ft3/*in)
Ci      =  concentration  of noble gas radionuclide        as determined by radioanalysis of grab sa*ple (uCi/c*3>
Ki      = total body dose conversion factor for noble gas radionucl i-de
                  <*re*/yr per uCi/*3, fro* Table 2-1>
Li      =  beta  skin dose conversion factor for noble gas radionuclide (nre*/yr per uCi/*3, from Table 2-1>
Hi      = gam*a air dose conversion factor for noble gas radionuclide (mrad/yr per uCi/m3, from Table 2-1) 1.1    = mrem skin dose per mrad gam*a air dose (mrem/mrad) 4.72E+02 = conversion factor <c*3/ft3
* min/sec>
500    = total body dose rate limit (mre*/yr) 3000    = skin dose rate li*it (*re*/yr)
* 15
Salem OOCWRev. 3 07/30/87 Based on the more limiting FRAC (i.e **      higher value) as determined  above. the alarm    setpoints    for the applicable monitors <Rl61  R41Cs  and/or R12A> may    be calculated by the equation:
SP  = CAF * '[Ci
* SEN + FRACJ + bkg                      (2.3>
where:
SP      = alarm setpoint corresponding to the maximum allowable release rate (Cp*)
SEN      = monitor sensitivity (cpm per uCi/c*3>
bkg      = background of the monitor (Cp*)
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 refulatory      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),
Creek.
will    be Any with the remainder 0.5 allocated to increase in AF above 0.5 for the Salem Nuclear Generating coordinated with the Hope Creek Generating Station to ensure that Hope Station the -
combined alloiation factors for all units do not exceed 1.0.
2.2.2                                        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    val~es  presented in Table 2-1 and 2-2 for Units 1 and 21    respectively *
* 16
Salem ODCH Rev. 3 07/30/87 These values are based upon:
* the maximu* ventilation Cor purge) flow rate:
a radionuclide distribution* comprised of 95X Xe-133, 2X Xe-135. 11 Xe-133** 11 Kr-88 and 11 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 maxi*um 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 Speci~ications. Table 6
* 17
                                                                  - Salem ODCH Rev. 3 07/30/87
* limits
  *re*/yr1 are the    dose    rate at the SITE BOUNDARY due to noble gas releaaes total body and    ~3000  mrem/yr, skin.
to  {500 Radiation monitor alar* aetpoints established to ensure that these release limits are not exceeded.            In    the event    any gaseous releases from the station results in an alarm aetpoint            being exceeded. an  evaluation    of  the SITE BOUNDARY doae rate    resulting  fro*    the release may be performed using the following equations:
Otb = X/Q *I:: CKi *Qi)                            <2.4) and Os  = X/Q
* l:_ CCLi + 1.1Hi)
* Qi)                (2.5) where:
Otb      =  total body dose rate <*re*/yr)
Da      =  skin dose rate <*re*/yr)
X/Q      =  at*ospheric dispersion to the controlling SITE BOUNDARY location c*eec/*3>
Qi      =  average release rate of radionuclide i over the release period under evaluation CuCi/sec>
Ki      =  total body dose conversion factor for noble gas radionuclide Cmre*/yr.per uCi/*31 fro* Table 2-1>
Li      =  beta skin dose conversion factor for noble gas radionuclide Cmre*/yr per uCi/*31 fro* Table 2-1>
Hi      =  gamma air dose conversion factor for noble gas radionuclide Cmrad/yr per uCi/*31 fro* Table 2-1) 1.1      =  11rem skin dose per 111rad gam*a air dose C111re*/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      3.11.2.1a,    following any release exceeding the above    prescribed alar*    setpointa.      Honitor    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 *onitor reading and monitor      sensitivity. The  15  *inute averaging    is    needed    to allow for reasonable monitor response      to  potentially changing radioactive *aterial concentrations and to exclude potential electronic
* 18-
Salem ODCH Rev. 3 07/30/87 spikes    in  monitor    readings    that may be  unrelated    to  radioactive    material releases
* As    identified,  any  electronic spiking monitor      responses  *BY    be
* excluded from the analysis.
NOTE:  For ad*inistrative purposes. more conservative alarm setpoints than those as prescribed 'above may be imposed.              However. conditions exceeding these more limiting alar* setpoints do not necessarily indicate radioactive material release rates exceeding the li*its of Technical Specification 3.11.2.1a.          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 de*onstrating compliance with the limits of Specification 3.11.2.1a.
Actual  meteorological    conditions    concurrent with the release      period  or    the default. annual    average dispersion parameters as presented in Table 2-4 *BY be used for evaluating the gaseous effluent dose rate.
2.3.2                                                                            Technical Specification 3.11.2.1.b limits the dose rate to          ~1500  mrem/yr to any organ      for 1-131,  tritium    and  particulates    with  half-lives greater      than  8  days. To demonstrate    Co*pliance  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 may be      used for the dose rate evaluation:
Oo  =  X/Q
* l:_  CRi
* Qi)                        (2.6) where:
Do    =  average organ doae rate over the sampling time period (mrem/yr)
X/Q    =  atmospheric dispersion to the controlling SITE BOUNDARY location for the inhalation. pathway <sec/*3>
Ri      =  dose parameter for radionuclide i1 (mrem/yr per uCi/*3> for the child inhalation pathway fro* Table 2-5 Qi      =  average release rate over the appropriate sampling period and analysis frequency for radionuclide i -- I-131, 1-1331 tritiu* or other radionuclide in particulate for* with half-life greater than 8 days <uCi/sec>
* 19
Salem ODCH Rev. 3 07/30/87 By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for  1-131  can  be  determined
* Based on the  annual  average  meteorological
* dispersion  <see Table 2-4> and the most limiting potential pathway, and organ (inhalation,    child, thyroid -- Ri allowable release rate for 1-131 is 42 uCi/sec.
                                                  = 1.62E+07 age 3
group mrem/yr per uCi/m ), the Reducing this release rate by a factor  of 4 to account for potential dose contributions from other    radioactive particulate  material  and  other  release  points *ce.g.,    Hope  Creek),    the corresponding release rate allocated to each of the Salem units is 10.S 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 1-131 releases in any 7 day period do not exceed 6.3 Ci1    no additional analyses are needed for verifying    compliance  with the Technical Specification    3.11.2.1.b  limits  on allowable release rate *
* 20
Salem ODCH Rev. 3 07130187
* 2.4.1 with the quarterly dose limits of        mrad, Technical Specification 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate compliance
                                        ~5        gamma-air and  ~10 mrad, beta-air and the calendar year limits    ~10 mrad, gamma-air and  ~20 mrad, beta-air. The limits
  ~re  applicable separately to each unit and    a~e  not combined site    limits. The following equations may be used to calculate the gamma-air and beta-air doses:
Dg    =  3.17E-08
* X/Q *[    <Hi
* Qi)                    (2.7) and Db    =. 3.17E-08
* X/Q
* C <Ni *Qi)                      (2.8) where:
Dg    =  air dose due to gamma emissions for noble gas radionuclides (mrad)
Db    =  air dose due to beta emissions for noble gas radionuclides (mrad)
X/Q    =  atmospheric dispersion to the controlling SITE BOUNDARY location
                  <sec/a3>
              =
Qi        cumulative release of noble gas radionuclide i over the period of interest CuCi)
Hi    = air dose factor due to gamma emissions from noble gas radionuclide i (arad/yr per uCi/m3, from Table 2-1>
Ni    =  air dose factor due to beta emissions from noble gas radionuclide (mrad/yr per uCi/m3, Table 2-1>
3.17E-08  =  conversion factor. (yr/sec>
In lieu of the    individual noble    gas  radionuclide    dose  assessment as  presented    above,  the  following simplified    dose  calculational equations may 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 aethod.>
* 21
Salem ODCH Rev. 3 07/30/87 3.17E-08 Og    =  --------
* X/Q
* HeH  * 'L  Qi                (2.9) a.so
* Ob  =
3.17E-08 a.so and
* X/Q
* X/Q
* VF* L CCi
* Neff * 't. Qi            <2.10) where:
* Ki)J + 500 <2.1> FRAC = C4.72E+02
He ff  = S.3E+02. effective ga1111a-air dose factor <mrad/yr per uCi/m3>
* X/Q *VF* [. CCi *<Li+ 1.1 Hi))J + 3000 <2.2> where: FRAC = X/Q = VF = Ci = Ki = Li = Hi = 1.1 = 4.72E+02 500 = 3000 = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate annual average meteorological dispersion to the controlling site boundary location (sec/m3) , ventilation system flow rate for the applicable release point and monitor (ft3/*in) concentration of noble gas radionuclide as determined by radioanalysis of grab sa*ple (uCi/c*3>
Neff  =  1.1E+03. effective beta-air dose factor (mrad/yr per uCi/m3>
total body dose conversion factor for noble gas radionucl i-de <*re*/yr per uCi/*3, fro* Table 2-1> beta skin dose conversion factor for noble gas radionuclide (nre*/yr per uCi/*3, from Table 2-1> gam*a air dose conversion factor for noble gas radionuclide (mrad/yr per uCi/m3, from Table 2-1) mrem skin dose per mrad gam*a air dose (mrem/mrad)
Qi    =   cumulative release for all noble gas radionuclides CuCi) a.so =     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 *
= conversion factor <c*3/ft3
* 22
* min/sec> total body dose rate limit (mre*/yr) skin dose rate li*it (*re*/yr) 15 
 
*
Salem ODCM Rev. 3 07/30/87 3.17E-08 Dg    = --------
* Salem OOCWRev. 3 07/30/87 Based on the more limiting FRAC (i.e ** higher value) as determined above. the alarm setpoints for the applicable monitors <Rl61 R41Cs and/or R12A> may be calculated by the equation:
where: SP = SEN = bkg = AF = SP = CAF * '[Ci
* SEN + FRACJ + bkg (2.3> alarm setpoint corresponding to the maximum allowable release rate (Cp*) monitor sensitivity (cpm per uCi/c*3> background of the monitor (Cp*) 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 refulatory 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 0.5 allocated to Hope Creek. Any increase in AF above 0.5 for the Salem Nuclear Generating Station will be coordinated with the Hope Creek Generating Station to ensure that the -combined alloiation factors for all units do not exceed 1.0. 2.2.2 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 presented in Table 2-1 and 2-2 for Units 1 and 21 respectively
* 16  
* *
* Salem ODCH Rev. 3 07/30/87 These values are based upon: the maximu* ventilation Cor purge) flow rate: a radionuclide distribution*
comprised of 95X Xe-133, 2X Xe-135. 11 Xe-133** 11 Kr-88 and 11 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 maxi*um 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 Table 6 17 
* * * -Salem ODCH Rev. 3 07/30/87 limits the dose rate at the SITE BOUNDARY due to noble gas releaaes to {500 *re*/yr1 total body and mrem/yr, skin. Radiation monitor alar* aetpoints are established to ensure that these release limits are not exceeded.
In the event any gaseous releases from the station results in an alarm aetpoint being exceeded.
an evaluation of the SITE BOUNDARY doae rate resulting fro* the release may be performed using the following equations:
Otb = X/Q *I:: CKi *Qi) <2.4) and Os = X/Q
* l:_ CCLi + 1.1Hi)
* Qi) (2.5) where: Otb = total body dose rate <*re*/yr)
Da = skin dose rate <*re*/yr)
X/Q = at*ospheric dispersion to the controlling SITE BOUNDARY location Qi Ki Li c*eec/*3>
= average release rate of radionuclide i over the release period under evaluation CuCi/sec>
= total body dose conversion factor for noble gas radionuclide Cmre*/yr.per uCi/*31 fro* Table 2-1> = beta skin dose conversion factor for noble gas radionuclide Cmre*/yr per uCi/*31 fro* Table 2-1> Hi = gamma air dose conversion factor for noble gas radionuclide Cmrad/yr per uCi/*31 fro* Table 2-1) 1.1 = 11rem skin dose per 111rad gam*a air dose C111re*/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 3.11.2.1a, following any release exceeding the above prescribed alar* setpointa.
Honitor 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 *onitor reading and monitor sensitivity.
The 15 *inute averaging is needed to allow for reasonable monitor response to potentially changing radioactive
*aterial concentrations and to exclude potential electronic 18-
* *
* Salem ODCH Rev. 3 07/30/87 spikes in monitor readings that may be unrelated to radioactive material releases
* As identified, any electronic spiking monitor responses
*BY be excluded from the analysis.
NOTE: For ad*inistrative purposes.
more conservative alarm setpoints than those as prescribed
'above may be imposed. However. conditions exceeding these more limiting alar* setpoints do not necessarily indicate radioactive material release rates exceeding the li*its of Technical Specification 3.11.2.1a.
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 de*onstrating compliance with the limits of Specification 3.11.2.1a.
Actual meteorological conditions concurrent with the release period or the default. annual average dispersion parameters as presented in Table 2-4 *BY be used for evaluating the gaseous effluent dose rate. 2.3.2 Technical Specification 3.11.2.1.b limits the dose rate to mrem/yr to any organ for 1-131, tritium and particulates with half-lives greater than 8 days. To demonstrate Co*pliance 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 may be used for the dose rate evaluation:
where: Do = X/Q = Ri = Qi = Oo = X/Q
* l:_ CRi
* Qi) (2.6) average organ doae rate over the sampling time period (mrem/yr) atmospheric dispersion to the controlling SITE BOUNDARY location for the inhalation.
pathway <sec/*3> dose parameter for radionuclide i1 (mrem/yr per uCi/*3> for the child inhalation pathway fro* Table 2-5 average release rate over the appropriate sampling period and analysis frequency for radionuclide i --I-131, 1-1331 tritiu* or other radionuclide in particulate for* with half-life greater than 8 days <uCi/sec>
19 
* *
* Salem ODCH Rev. 3 07/30/87 By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for 1-131 can be determined
* Based on the annual average meteorological dispersion
<see Table 2-4> and the most limiting potential pathway, age group 3 and organ (inhalation, child, thyroid --Ri = 1.62E+07 mrem/yr per uCi/m ), the allowable release rate for 1-131 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 *ce.g., Hope Creek), the corresponding release rate allocated to each of the Salem units is 10.S 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 1-131 releases in any 7 day period do not exceed 6.3 Ci1 no additional analyses are needed for verifying compliance with the Technical Specification 3.11.2.1.b limits on allowable release rate
* 20 
* * * -Salem ODCH Rev. 3 07130187 2.4.1 Technical Specification 3.11.2.2 requires a periodic assessment of releases of noble gases to evaluate compliance with the quarterly dose limits of mrad, gamma-air and mrad, beta-air and the calendar year limits mrad, gamma-air and mrad, beta-air.
The limits applicable separately to each unit and not combined site limits. The following equations may be used to calculate the gamma-air and beta-air doses: Dg = 3.17E-08
* X/Q * [ <Hi
* Qi) (2.7) and Db =. 3.17E-08
* X/Q
* X/Q
* C <Ni *Qi) (2.8) where: Dg = air dose due to gamma emissions for noble gas radionuclides (mrad) Db = air dose due to beta emissions for noble gas radionuclides (mrad) X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location Qi Hi Ni <sec/a3> = cumulative release of noble gas radionuclide i over the period of interest CuCi) = air dose factor due to gamma emissions from noble gas radionuclide i (arad/yr per uCi/m3, from Table 2-1> = air dose factor due to beta emissions from noble gas radionuclide (mrad/yr per uCi/m3, Table 2-1> 3.17E-08 = conversion factor. (yr/sec> In lieu of the individual noble gas radionuclide dose assessment as presented above, the following simplified dose calculational equations may 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 aethod.> 21 
* *
* where: He ff = Neff = Qi = a.so = Salem ODCH Rev. 3 07/30/87 3.17E-08 Og = --------* X/Q
* HeH * 'L Qi (2.9) a.so and 3.17E-08 Ob = --------* X/Q
* Neff * 't. Qi a.so <2.10) S.3E+02. effective ga1111a-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 CuCi) 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
* 22 
* *
* where: He ff = Neff = Qi = o.so = Dg Db Salem ODCM Rev. 3 07/30/87 3.17E-08 = --------* X/Q
* Heff
* Heff
* I: Qi a.so and 3.17E-08 = --------* X/Q
* I: Qi               (2.9) a.so
* Db  = --------
a.so and 3.17E-08
* X/Q
* Neff
* Neff
* t_ Qi a.so (2.9) <2.10) -s.3E+021 effective gamma-air dose factor (mrad/yr per 1.1E+Q3, effective beta-air dose factor (mrad/yr per uCi/m > cumulative release for all noble gas radionuclidea CuCi> conservatism factor to account for potential variability in the radionuclide distribution Actual meteorological condition*
* t_ Qi           <2.10) where:
concurrent with the release period or the default, annual average dispersion parameters aa presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doaea
He ff  = -s.3E+021 effective gamma-air dose factor (mrad/yr per uCi/~ 3 >
* 22
Neff  = 1.1E+Q3, effective beta-air dose factor (mrad/yr per uCi/m >
* *
Qi    = cumulative release for all noble gas radionuclidea CuCi>
* Salem ODCH Rev. 3 07/30/87 2.5 Radioiodine and Particulate Dose Calculations  
o.so  = conservatism factor to account for potential variability in       the radionuclide distribution Actual   meteorological   condition* concurrent with the release period or   the default,   annual average dispersion parameters aa presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doaea *
-10 CFR 50 2.5.1 UNRESTRICTED AREA Ooae -Radioiodine and Particulates.
* 22
In accordance with requirements of Technical Specification 3.11.2.31 a periodic aeaeasment shall be perforaed to evaluate compliance with the quarterly dose limit of i7.5 mrem and calendar year limit i15 mrem to.any organ. The following equation may be used to evaluate the maximum organ dose due to releases of I-1311 tritium and particulates with half-lives greater than 8 days: Daop = 3.17E-08
 
Salem ODCH Rev. 3 07/30/87 2.5   Radioiodine and Particulate Dose Calculations - 10 CFR 50
* 2.5.1 with shall UNRESTRICTED AREA Ooae - Radioiodine and Particulates.
requirements of Technical Specification 3.11.2.31         a In periodic accordance aeaeasment be perforaed to evaluate compliance with the quarterly dose limit of i7.5 mrem and calendar year limit i15 mrem to.any organ.           The following equation may be used to evaluate the maximum organ dose due to releases of I-1311 tritium and particulates with half-lives greater than 8 days:
Daop   = 3.17E-08
* W
* W
* SFp
* SFp
* L. CRi
* L. CRi
* Qi> <2.11) where: Daop = dose or dose commitment via controlling pathway p and age group a <a* identified in Table 2-4> to organ o. including the total body <111rem) Ri = atmospheric dispersion parameter to the controlling location(e) as identified in Table 2-4 = X/Q = atmospheric dispersion for inhalation  
* Qi>             <2.11) where:
!athway and dose contribution via other pathways Csec/m ) D/Q = atmospheric deposition milk and ground plane exposure pathways Cm 3 2 dose factor for radionuclide i1 <mrem/yr per uCi/m) or (m mrem/yr per uCi/mec> from Table 2-5 for each age group a and the applicable pathway pas identified in Table 2-4. Values for R. were derived in accordance with the methods described in NUREG! 0133. Qi = cumulative releaee over the period of interest for radionuclide i --I-131 or radioactive material in particulate form with half-1 ife greater than 8 days CuCi). SFp = annual *eaeonal correction factor to account for the fraction of the year that the applicable expo*ure pathway does not exist. 1> For *ilk and vegetation exposure pathways:  
Daop   = dose or dose commitment via controlling pathway p and age group a
= A eix month fresh vegetation and grazing season <Hay through October) = o.s 2) For inhalation and ground plane exposure pathways:  
                  <a* identified in Table 2-4> to organ o. including the total body
= 1.0 For evaluating the 111axi111um exposed individual 1 the infant age group is controlling for the milk pathway and the child age group is controlling for the 23 Salem ODCH Rev. 3 07/30/87 vegetation pathway. Only the controlling age group and pathway aa identified in
                  <111rem)
* Table 2-4 need be evaluated for compliance with Technical Specification  
              = atmospheric dispersion parameter to the controlling location(e) as identified in Table 2-4 X/Q   = atmospheric dispersion for inhalation !athway and H-~ dose contribution via other pathways Csec/m )
*
D/Q   = atmospheric deposition fo~ 2 vegetation1 milk and ground plane exposure pathways Cm 3 or (m 2 Ri    =  dose factor for radionuclide i1 <mrem/yr per uCi/m) mrem/yr per uCi/mec> from Table 2-5 for each age group a and the applicable pathway pas identified in Table 2-4.             Values for R.
* 3.11.2.3.
were derived in accordance with the methods described in NUREG!
2.5.2 Simplified Dose Calculation for Radioiodinea and Particulatea.
0133.
In lieu of the individual radionuclide CI-131 and particulates>
Qi     = cumulative releaee over the period of interest for radionuclide i
doae aaseaament aa presented above. the following simplified doae calculational equation may be uaed for -verifying compliance with the doae limita of Technical Specification 3.11.2.3 (refer to Appendix D for the derivation and justification of thia simplified method). where: Dmax * = RI-131 = = w = Qi = Dmax = 3.17E-08
                  -- I-131 or radioactive material in particulate form with half-1 ife greater than 8 days CuCi).
SFp   = annual *eaeonal correction factor to account for the fraction of the year that the applicable expo*ure pathway does not exist.
1> For *ilk and vegetation exposure pathways:
                        = A eix month fresh vegetation and grazing season <Hay through October)
                        = o.s
: 2)   For inhalation and ground plane exposure pathways:
                        = 1.0 For   evaluating   the   111axi111um exposed individual 1 the   infant age group is controlling for the milk pathway and the child age group is controlling for           the
* 23
 
Salem ODCH Rev. 3 07/30/87 vegetation pathway.       Only the controlling age group and pathway aa identified in
* Table   2-4   need   be   evaluated   for   compliance   with   Technical   Specification 3.11.2.3.
2.5.2     Simplified Dose Calculation for Radioiodinea and Particulatea.               In lieu of   the   individual     radionuclide CI-131 and particulates>         doae aaseaament   aa presented     above. the following simplified doae calculational equation           may   be uaed   for -verifying compliance with the doae limita of Technical             Specification 3.11.2.3     (refer     to Appendix D for         the     derivation     and justification of thia simplified method).
Dmax = 3.17E-08
* W
* W
* SFp
* SFp
* RI-131 .-[_Qi <2.12) maximum organ dose <*rem> I-131 dose parameter for the thyroid for the identified controlling pathway 1.05E+121 thyroid doae parameter with the cow-milk pathway controlling (m -mrem/yr per uCi/eec) D/Q for radioiodine1 2.lE-10 1/* cumulative releaae over the period of interest for radionuclide i --I-131 or radioactive material in particulate form with half life greater than 8 daya (uCi) The location of expoeure pathways and the maximum organ do*e calculation may be based on the available pathway* in the surrounding environ*ent of Salem as identified by the annual land-u*e census <Technical Specification 3.12.2>. Otherwise, the doae will be baaed on the 9redetermined controlling pathwaya as identified in Table
* RI-131 .- [_Qi             <2.12) where:
* 24
Dmax *  = maximum organ dose <*rem>
* *
RI-131  = I-131 dose parameter for the thyroid for the identified controlling pathway
* Salem ODCH Rev. 3 07/30/87 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.
                  = controlling 1.05E+121 infa~t thyroid doae parameter with the cow-milk pathway (m - mrem/yr per uCi/eec) w        =  D/Q for radioiodine1   2.lE-10 1/*
Non-condensables
Qi      =  cumulative   releaae   over the period of interest for radionuclide i -- I-131 or radioactive material in particulate form with half life greater than 8 daya (uCi)
<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 sa*pling syetem and procedures (e.g., R15 on condenser evacuation.
The   location of expoeure pathways and the maximum organ do*e calculation may be based   on   the   available pathway* in the surrounding environ*ent           of   Salem   as identified     by   the   annual land-u*e census     <Technical     Specification   3.12.2>.
R41C on plant vent. and the plant vent particulate and charcoal samplers).
Otherwise,     the   doae will be eva~uated    baaed on the     9redetermined   controlling pathwaya as identified in Table       2~4 *
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 Ci.e ** releases due to moisture carry over). Since this pathway is not sampled or monitored.
* 24
it is necessary to calculate potential releases
 
* Based on the guidance in NRC NUREG-01331 the releases of the radioiodines and particulates may be calculated by the equation:
Salem ODCH Rev. 3 07/30/87 2.6   Secondary Side Radioactive Gaseous Effluents and Dose Calculations
Qi = Ci
* During material periods may condensables be of primary to secondary leakage,
                  <e.g.,     noble   gases) will   be minor levels released via the secondary system to    the predominately of  radioactive atmosphere.
released   via Non-the condenser evacuation system and will be monitored and quantified by the           routine plant vent Monitoring and sa*pling syetem 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   Ci.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-01331       the releases of the radioiodines     and particulates may be calculated by the equation:
Qi   = Ci
* Rsgb
* Rsgb
* Fft
* Fft
* C1-SQftv>  
* C1-SQftv>                   <Z.13>
<Z.13> where: Qi = the release rate of radionuclide.
where:
i1 from the steaa generator flash Ci = Rsgb = Fft = SQftv = tank vent CuCi/aec>
Qi     = the release rate of radionuclide. i1 from the steaa generator               flash tank vent CuCi/aec>
the concentration of radionuclide.
Ci    =  the concentration of radionuclide. i1 in the secondary coolant             water averaged over not *ore than one week CuCi/ml>
i1 in the secondary coolant water averaged over not *ore than one week CuCi/ml> the stea* generator blowdown rate to the flash tank (ml/sec> the fraction of blowdown flashed in the tank determined fro* a heat balance taken around the flash tank at the applicable reactor power level the measured steam quality in the flash tank vent: or an assumed value of 0.851 based on NUREG-0017
Rsgb  =  the stea* generator blowdown rate to the flash tank (ml/sec>
* 25
Fft    =  the fraction of blowdown flashed in the tank determined fro* 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 0.851 based on NUREG-0017 *
* * *** Salem ODCH Rev. 3 07/30/87 Tritium releases via the steam flashing may also be quantified using the above equation with the aseumption of a steam quality <SQftv) equal to O. Since the H-3 will be associated with the water molecules.
* 25
it is not necessary to account for the moisture carryover which is the transport media for the radioiodinea and particulates.
 
Based on the design and operating conditions at Sale** the fraction of blowdown converted to steam <Fft> is approximately 0.48. following:
Salem ODCH Rev. 3 07/30/87 Tritium   releases via the steam flashing may also be quantified using the           above
Qi = 0.072 Ci Rsgb For H-3, the simplified equation is! Qi = 0.48 Ci Rsgb The equation simplifies to the <2.14) (2.15) Also during reactor shutdown operations with a radioactively contaminated secondary system. radioactive material may be released to the atmosphere via the atmospheric reliefs <PORV> and the safety reliefs on the main steam lines and 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-00171 Revision 1. The partitioning factors for the radioiodines is 0.01 and is 0.001 for all other particulate radioactive Material.
* equation with the aseumption of a steam quality <SQftv) equal to O. Since the H-3   will be associated with the water molecules.       it is not necessary to for the moisture carryover which is the transport media for the radioiodinea and particulates.
The resulting equation for quantifying releases via the atmo*pheric stea* releases is: 26 
account Based on the design and operating conditions at Sale**         the fraction of blowdown converted to steam <Fft> is approximately 0.48.         The equation simplifies to the following:
* *
Qi   = 0.072 Ci   Rsgb                             <2.14)
* where: Qij SFj PF Salem ODCH Rev. 3 07/30/87 Qi = 0.13
For H-3, the simplified equation is!
Qi = 0.48 Ci   Rsgb                             (2.15)
Also   during   reactor   shutdown   operations with   a radioactively   contaminated
* secondary system. radioactive material may be released to the atmosphere via the atmospheric vi~  the reliefs <PORV> and the safety reliefs on the main steam steam   driven   auxiliary feed pump exhaust.       The evaluation lines of and the radioactive   material concentration in the steam relative to that in         the   steam generator   water   is based on the guidance     of   NUREG-00171   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 atmo*pheric stea* releases is:
***                                    26
 
Salem ODCH Rev. 3 07/30/87 l
Qi = 0.13
* L_ CCij
* L_ CCij
* SFj) *PF =release rate of radionuclide i via pathway j <uCi/sec)  
* SFj) *PF                                   (2.16) where:
= steam flow for release pathway j = 4501000 lb/hr per PORV = 8001000 lb/hr per safety relief valve = 501000 lb/hr for auxiliary feed pump exhaust = partitioning factor, ratio of concentration in steam to that in the water in the steam generator  
Qij  =release rate of radionuclide i via pathway j <uCi/sec)
= 0.01 for radioiodines  
SFj  = steam flow for release pathway j
= 0.005 for all other particulates  
            = 4501000 lb/hr per PORV
= 1.0 for H-3 (2.16) 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.
            = 8001000 lb/hr per safety relief valve
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.
            = 501000 lb/hr for auxiliary feed pump exhaust PF  = partitioning factor, ratio of concentration in steam to that in the water in the steam generator
The measured radionuclide concentration in the steam may be used for quantifying the noble gases, radioiodine and particulate releases.
            = 0.01 for radioiodines
Note: The expected mode of operation wouid be to isolate the effected steam generator.
            = 0.005 for all other particulates
thereby reducing the potential releases during the shutdown/cooldown process. Use of the above calculational methods should consider actual operating conditions and release mechanisms.
            = 1.0 for H-3 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.
The calculated quantities of radioactive materials may be used as inputs to the equation <2.11> or C2.12) to calculate offsite doses for demonstrating compliance with the Radiological Effluent Technical Specifications
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.
* 27 l 
The measured radionuclide concentration in the steam may be used for quantifying
* *
* the noble gases, radioiodine and particulate releases.
* Salem ODCH Rev. 3 07/30/87 2.7 Gaaeoua Effluent Doae Projection Technical Specification 3.11.2.4 requires that the GASEOUS RADWASTE TREATHENT SYSTEM and VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce radioactive material levels prior to discharge when projected dosee exceed one-half the annual deaign objective rate in any calendar quarter, i.e ** exceeding:
Note: The expected mode of operation wouid be to isolate the effected steam generator. thereby reducing the potential releases during the shutdown/cooldown process.     Use of the above calculational methods should consider actual operating conditions and release mechanisms.
0.625 mrad/quarter.
The calculated quantities of radioactive materials may be used as inputs to     the equation   <2.11>   or C2.12)   to calculate offsite doses   for demonstrating compliance with the Radiological Effluent Technical Specifications *
gamma air: 1.25 mrad/quarter, beta air: or  
* 27
*re*/quarter.
 
maximum organ. The applicable gaseous proceeeing ayete*a for *aintaining radioactive material releaaee ALARA are the Auxiliary Building normal ventilation eyatem (filtration syete*a M 112 and 3) and the Waate Gaa Decay Tanke ae delineated in Figures 2-3 and 2-4. Doee projection&
Salem ODCH Rev. 3 07/30/87 2.7 Gaaeoua Effluent Doae Projection
are perfor*ed at least once per 31 days by the following equationa:
* Technical SYSTEM material and Specification 3.11.2.4 requires that the GASEOUS   RADWASTE VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce levels   prior to discharge when projected dosee   exceed TREATHENT radioactive one-half   the annual deaign objective rate in any calendar quarter, i.e ** exceeding:
D gp = 09 * <91+ d) (2.17> D bp = Db * (91+ d> (2.18) Dmaxp = D*ax * (91+ d> <2.19) where: D gp = Og D bp = Db = Dmaxp = D*ax = d = 91 = = 9am*a air doee projection for current calendar quarter Cmrad) 9a**a air doae to date for current calendar quarter aa determined by equation <2.7> or <2.9) <*rad) beta air doae projection for current calendar quarter <*rad) beta air doee to date for current calendar quarter ae determined by equation <2.8) or <2.10) (mrad> maximu* organ doae projection for current calendar quarter (mrem> maximu* organ doae to date for current calendar quarter aa determined by equation <2.11) or <2.12> (mre*> nu*ber of daya to date in current calendar quarter number of daya in a calendar quarter 28 Salem OOCH Rev. 3 07/30/87 3.0 Special Dose Analyses
0.625 mrad/quarter. gamma air:
* 3.1 Doses Due To Activities Inside the SITE BOUNDARY In accordance with Technical Specification 6.9.1.11.
1.25 mrad/quarter, beta air: or
the Radioactive Effluent Re)ease Report <RERR> submitted within 60 days after January 1 of each year shall include an aseeesment of radiation doses fro* radioactive liquid and gaseoue effluent*
          ~.875 *re*/quarter. maximum organ.
to HEHBERS OF THE PUBLIC due to their activities ineide the SITE BOUNDARY.
The applicable gaseous proceeeing ayete*a for *aintaining radioactive     material releaaee ALARA are the Auxiliary Building normal ventilation eyatem     (filtration syete*a   M 112 and 3) and the Waate Gaa Decay Tanke ae delineated in Figures 2-3 and 2-4.
There is one location on Artificial Island that is acceesible to HEHBERS OF THE PUBLIC for unrelated to PSE&G operational and support activitiea.
Doee   projection&   are perfor*ed at least once per 31 days   by the   following equationa:
This location i* the Second Sun <vieitor*s center> located near the contractors gate for the Salem Nuclear Generating Station. The calculation Methods ae presented in Section* 2.4 and 2.5 may be used for determining the maxi*um potential dose to a HEHBER OF THE PUBLIC based on the
* D gp D bp Dmaxp
* para*eters fro* Table 2-4 and 2 houre per visit per year. The default value for the mereorological dispereion data as presented in Table 2-3 *BY be used if current year *eteorology in unavailable at the ti*e of NRC reportin.
                = 09 * <91+ d)
However. a follow-up evaluation shall be performed when the data beco*ee available  
                = Db * (91+ d>
* * . 29
                = D*ax * (91+ d>
* *
(2.17>
* Salem ODCH Rev. 3 07/30/87 3.2 Total dose to HEHBERS OF THE PUBLIC -40 CFR 190 The Radioactive Effluent Releaee Report <RERR> submitted within 60 days after January 1 of each year ahall aleo include an aseess*ent of ihe radiation doae to the likely moat exposed HEHBER OF THE PUBLIC for reactor releaeea and other nearby uranium fuel cycle sources (including doae contributions from effluents and direct radiation from on-site sources>.
(2.18)
For the likely most expoeed HEHBER OF THE-PUBLIC in the vicinity of Artificial Ieland1 the source* of exposure need only consider the Sale* Nuclear Generating Station and the Hope Creek Nuclear Generating Station: No other fuel cycle facilities contribute to the HEHBER OF THE PUBLIC doae for the Artificial Ieland vicinity.
                                                                                <2.19) where:
The doee contribution from the operation of Hope Creek Nuclear Generating Station will be estimated baaed on the *ethoda as presented in the Hope Creek Offaite Dose Calculation Hanual CHCGS ODCH>
D gp     = 9am*a air doee projection for current calendar quarter Cmrad)
* As appropriate for de*onstrating/evaluating compliance with the limits of Technical Specification 3.11.4 <40 CFR 190)1 the result* of the environmental monitoring program *ay be used for providing data on actual measured levels of radioactive material in the actual pathways of exposure.
Og       = 9a**a air doae to date for current calendar quarter aa determined by equation <2.7> or <2.9) <*rad)
3.2.1 Effluent Doee Calculation*.
D bp    = beta air doae projection for current calendar quarter <*rad)
For purpoees of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.11 CRERR>s doae calculations for the Salem Nuclear Generating Station may be performed using the calculational method* contained within thie ODCH: the conservative controlling pathway* and locations of Table 2-4 or the actual pathway* and locations ae identified by the land use ceneue <Technical Specification 3/4.12.2) may be uaed. Average annual meteorological diapereion 30
Db      = beta air doee to date for current calendar quarter ae determined by equation <2.8) or <2.10) (mrad>
* *
Dmaxp    = maximu* organ doae projection for current calendar quarter (mrem>
* Salem ODCH Rev. 3 07/30/87 para*eters or *eteorological condition*
D*ax    = maximu* organ doae to date for current calendar quarter aa determined by equation <2.11) or <2.12> (mre*>
concurrent with the release period under evaluation may be used
d      = nu*ber of daya to date in current calendar quarter 91      = number of daya in a calendar quarter
* 3.2.2 Direct Exposure Pose Determination.
* 28
Any potentially significant direct exposure contribution to off-site individual doees may be evaluated based on the results of the environmental  
 
*easurementa  
Salem OOCH Rev. 3 07/30/87 3.0   Special Dose Analyses
<e.g.. TLD1 Ion cha*ber meaaurements>
* 3.1 In Doses Due To Activities Inside the SITE BOUNDARY accordance with Technical Specification 6.9.1.11.
and/or by the uee of a radiation traneport and ehielding calculational method. Only during atypical condition*
Re)ease   Report <RERR>
will there exiet any potential for significant on-site eources at Sale* that would yield potentially significant off-site doees Ci.e., in excess of 1 mrem per year to a HEHBER OF THE PUBLIC>. that would require detailed evaluation for de*onstrating compliance with 40 CFR 190. However. should a situation exist whereby the direct exposure contribution is potentially significant.
the Radioactive submitted within 60 days after January 1 of Effluent each   year shall   include an aseeesment of radiation doses fro*     radioactive liquid   and gaseoue   effluent* to HEHBERS OF THE PUBLIC due to their activities     ineide   the SITE BOUNDARY.
on-site measure*ente.
There   is one location on Artificial Island that is acceesible to HEHBERS OF THE PUBLIC   for activi~ies  unrelated to PSE&G operational and     support activitiea.
off-site meaeurements and/or calculational techniques will be used for determination of dose for aseesaing 40 CFR 190 compliance
This location i* the Second Sun <vieitor*s center> located near the contractors gate for the Salem Nuclear Generating Station.
* 31
The   calculation Methods ae presented in Section* 2.4 and 2.5 may be used       for determining   the maxi*um potential dose to a HEHBER OF THE PUBLIC based     on   the
*
* para*eters fro* Table 2-4 and 2 houre per visit per year.       The default value for the   mereorological   dispereion   data as presented in Table 2-3 *BY be   used   if current year *eteorology in unavailable at the ti*e of NRC reportin.       However. a follow-up evaluation shall be performed when the data beco*ee available *
* Salem ODCH Rev. 3 07/30/87 4.D Radiological Environ*ental Monitoring Program 4.1 Sa*pling Program The operational phaae of the Radiological Environmental Horiitorin9 Progra11 CREHP> ia conducted in accordance with the require11enta of Appendix A Technical Spec if icat ion 3 .12. The object i vee of the pro9ra11 are: -To determine whether *ny ei9nificant increaaea occur in the concentration of radionuclide*
*                                 . 29
in the critical pathway* of exposure in the vicinity of Artificial Ieland: -To deter11ine if the operation of the Salem Nuclear Generatin9 Stations ha* reeulted in any increase in the inventory of lon9 lived radionuclidea in the To detect any changes in the ambient ga1111a radiation levela: and -To verify that SNGS operation*
 
have no detrimental effecta on the health and eafety of the pub 1 ic or on the environ*ent.
Salem ODCH Rev. 3 07/30/87 3.2   Total dose to HEHBERS OF THE PUBLIC - 40 CFR 190 The   Radioactive   Effluent Releaee Report <RERR> submitted within 60 days       after January 1 of each year ahall aleo include an aseess*ent of ihe radiation doae to the   likely   moat exposed HEHBER OF THE PUBLIC for reactor     releaeea   and   other nearby   uranium fuel cycle sources (including doae contributions from         effluents and direct radiation from on-site sources>.         For the likely most expoeed HEHBER OF THE- PUBLIC in the vicinity of Artificial Ieland1 the source* of exposure need only   consider   the Sale* Nuclear Generating Station and the Hope Creek       Nuclear Generating Station:     No other fuel cycle facilities contribute to the HEHBER OF THE PUBLIC doae for the Artificial Ieland vicinity.
The aa11plin9 require111enta  
The   doee   contribution   from the operation of   Hope Creek   Nuclear   Generating Station   will   be estimated baaed on the *ethoda as presented in the Hope       Creek Offaite Dose Calculation Hanual CHCGS ODCH>
<type of sa*ples, collection frequency and analysis) and sa*ple location*
* As   appropriate   for de*onstrating/evaluating compliance     with the   limits   of Technical   Specification 3.11.4 <40 CFR 190)1     the result* of the   environmental monitoring   program *ay be used for providing data on actual measured levels         of radioactive material in the actual pathways of exposure.
are presented in Appendix E
3.2.1 Effluent Doee Calculation*.       For purpoees of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting         requirements of   6.9.1.11 CRERR>s   doae calculations for the Salem Nuclear Generating Station may be performed using the calculational method* contained within thie ODCH: the conservative     controlling   pathway*   and locations of Table 2-4   or   the   actual pathway*   and   locations   ae   identified by the land use   ceneue   <Technical Specification 3/4.12.2) may be uaed.         Average annual meteorological   diapereion
* 32
* 30
* *
 
* Salem OOCH Rev. 3 07/30/87 4.2 Interlaboratorv Comparison Proqra* Technical Specification 3.12.3 requires analyses be perforaed on radioactive material supplied as part of an Interlaboratory Comparison.
Salem ODCH Rev. 3 07/30/87 para*eters or *eteorological condition* concurrent with the release period under evaluation   may be used *
Participation in an approved Interlaboratory Comparison Prograa provides a check on the precisness 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
* 3.2.2 Direct   Exposure Pose Determination. Any potentially significant exposure contribution to off-site individual doees may be evaluated based on the direct results of the environmental *easurementa <e.g.. TLD1 Ion cha*ber meaaurements>
and/or   by the uee of a radiation traneport and ehielding calculational       method.
Only   during atypical condition* will there exiet any potential for       significant on-site eources at Sale* that would yield potentially significant off-site doees Ci.e.,   in   excess of 1 mrem per year to a HEHBER OF THE   PUBLIC>. that   would require   detailed   evaluation for de*onstrating compliance   with   40   CFR   190.
However. should a situation exist whereby the direct exposure contribution       is potentially   significant. on-site measure*ente. off-site meaeurements     and/or calculational techniques will be used for determination of dose for aseesaing 40 CFR 190 compliance *
* 31
 
Salem ODCH Rev. 3 07/30/87 4.D   Radiological Environ*ental Monitoring Program 4.1 Sa*pling Program
* The CREHP>
operational     phaae of the Radiological   Environmental   Horiitorin9   Progra11 ia conducted in accordance with the require11enta of Appendix A Technical Spec if icat ion 3 .12. The object i vee of the pro9ra11 are:
        - To determine whether *ny ei9nificant increaaea occur in the concentration of radionuclide* in the critical pathway* of exposure in the vicinity of Artificial Ieland:
        - To deter11ine i f the operation of the Salem Nuclear Generatin9 Stations ha* reeulted in any increase in the inventory of lon9 lived radionuclidea in the environ~ent:
To detect any changes in the ambient ga1111a radiation levela: and
        - To verify that SNGS operation* have no detrimental effecta on the health and eafety of the pub 1 ic or on the environ*ent.
The aa11plin9 require111enta <type of sa*ples,     collection frequency and     analysis) and sa*ple location* are presented in Appendix E*
* 32
 
Salem OOCH Rev. 3 07/30/87 4.2 Interlaboratorv Comparison Proqra*
* Technical   Specification 3.12.3 requires analyses be perforaed material supplied as part of an Interlaboratory Comparison.
on  radioactive Participation in an approved Interlaboratory Comparison Prograa provides a check on the   precisness 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 *
* 33
* 33
* ro -If.I FUER iii RADIATION MONITOR &#xa3;::, INSTRUMENT PANEL -#-PNEUMATIC -ELECTiaC.AL
 
-+--UOUID EFFLUENT -+-GAS EFFLUENT ... .. ... ... ... ... *M l<i
                                                            ----t...... rn'I HU.1,.:100 COlJl.AHI
* HU.1,.:100 COlJl.AHI SEAL WAIUC ----t ... -l(Il FltlER , St:A.1 WAIEH ... rn'I --ft *
                                                ~                    -l(Il SEAL WAIUC FltlER ,
---1=, sv nMS1J1 17MSl:l1 -* ** COMP ID "" 10 .... ....... ...,,,.. ... -....... .... --.... -** --....... ..... .. *** -....... .... -ACI011A ... -...,, .. -ffAI ... ...,, .. ... RAIUI -.... ffA1"'9 ....... -ffAI-...... .... -.... -.... Rw111 -.... --** --....... .... ....,., _,. ... --....... ---""' ....... -... RIWlll -llU ....,.. ........ ... -... ....... ... ------....... -....... "'"" ...... ....., ...... --* FOR INFORMATION ONLY FIGURE 1-1 RADIATION MONITORING UOUID 1UNIT 'I I' I 
                                                                                        --ft St:A.1 WAIEH
** lfJIJ()WH  
                                                                                                                            ~
... TO OMNllCI ------1 -. --1
ID COMP 10 ACI011A ffAI...
__ : : = t= .. ----L---* fAIWIO lEOENp !El flJ'&I Ii RADIATION lllOHITOR 6 INSDIUIENT MHEL _,,_.PNEUllATIC -EU:CIRICAL  
RAIUI ffA1"'9 ffAI-Rw111 RIWlll llU                    ....,.. ........
-+-IJOUI> EfR.UENT -+-GAS EfR.UENT
~
* 121 <XlHlllNSER CtlNdlHSlff ... llMSlll 11t&#xa3;11t1 .. ntilS1l1 ........ ... .. '"' -.. 111 -..
ro
                                                                        *1r---::--~~H'!.----------r-- ---1=,
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sv
                                                                                                                                                                        'I I
FOR INFORMATION ONLY If.I iii FUER RADIATION MONITOR
                            ....       nMS1J1 FIGURE 1-1
    &#xa3;::, INSTRUMENT PANEL              17MSl:l1 RADIATION
    -#-PNEUMATIC
  -ELECTiaC.AL              ......                                                                           MONITORING UOUID
  -+-- UOUID EFFLUENT
  -+-GAS EFFLUENT
                            *M l<i    ~JD 1UNIT
 
                                                                                                                        -1.IL.
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BtlC lfJIJ()WH AA1Rr
                                                                                                                            --- --          -tO
  -TO OMNllCI
                          - - ---1
              +-"=::Bi--~* - . --1                                                     121 SUIVIC&#xa3;    121XN>
  ~::*~-;;,~-~*-:___ : : =~~1 lllllAICA~A      alC WIR OliDi 121 <XlHlllNSER t=.. -
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        !El flJ'&I                 ..... llMSlll 11t&#xa3;11t1 FIGURE 1-2 Ii RADIATION lllOHITOR 6 INSDIUIENT MHEL ntilS1l1
                                                          ........
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      -EU:CIRICAL                   ..
                                    '"'                                              *nSUNU
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MONITORING UOUID
      -+- IJOUI> EfR.UENT
      -+-GAS EfR.UENT               ..
111 2UNIT
* Monitoring for the 124 and 125 containment fan coil units is provided by the Rl3A, B or C monitors through interconnects.
* Monitoring for the 124 and 125 containment fan coil units is provided by the Rl3A, B or C monitors through interconnects.
121 SUIVIC&#xa3;
* WU HUT OTHER UNIT FROM LAUNDRY & CHEM TKS              2UNIT                EOUIPORS WHUTS                AUX BlllG SU ... TK                                          RCPHEADTKS FLOOR DRS                                                    RC LOOP ORS Wl68      Wl172                          Cl/CS HUT RELIEFS                                            RX FLANGE LO Wl.177                                                                                              RHRSUMl'S                                                  ACCUMUUUOR ORS EXCESS LET DOWN FROM llAEVN' Wl70 RWPP I                                                                  Wl222 I                                                      RWST I                      1WHUT            Wl71                    WI.Ill:!
*nSUNU *AIOI HEADBI * .,..._, PStiGCO -1.IL. Mlil"'1. .... .. .... .... ..,... BtlC .. _ .. ,. ...,,. .... ....,.. ... ...,,. .... .. ... -AA1Rr ... ... ... -....... .... --... .... ... , ----.. , -------tO -....... ----------... -* --* ...., ....... ----121XN> alC WIR OliDi
RCPS 1 VVMT      Wl 8ti 2VVMT
* FOR INFORMATION
                                                          "'              Wl71 CllCSHUT                                Wl.7      RCDT              3SEALLO I
TO&FROM OTHER UNIT Wl..84 2WHUT FC COHTSUMP Wl73 WASTEEVAP
                                                                                                  ~-  ...I FEEOPUMP        Wl.178 WASTEEVAP:
Wl.11                          I I
Wl.47
                                                                                                        'f'                      2Wl.174      2Wl.193
                                                                  -------~-------~------~---w-+--W-,-"*'--4~
                                                                                                        ~
WASTE                        Wl.11 I                                                                                                                                                                EVAP I                                                                                                                                                                                            Wl.77 I
I I                                                                                                                                                                                    Wl.1711 I
L-----                                                                                                                                            OllERllOARO ORWMTS Wl.183        2WMHUT Wl.115 TOANDFROM                                                                                                                                        PUMP IWi.2611        1---ciO-+ OTHER UNIT                                                                                                            12 2lWl6 -
FOR INFORMATION ONLY II SW222 ORC WAJER DISCH 21 SERVICE WAJER                                                                                      -~*"--r~~                                                                              FIGURE 1-3 LIQUID I.WI&
WASTE ILW41
 
Sale1 ODc"* Rev. 3 07/30/87 Table 1-1 Para1eters for
* VOL
* VOL
* A Nb-95,GI-LLI Dmax = ---------------------------
* A Nb-95,GI-LLI Dmax   = ---------------------------
* L: Ci cw = maximum organ dose (mrem) <B.2> <B.3> = 1.51E+06, Gi-LLI ingestion dose conversion factor for Nb-95 Cmrem/hr per uCi/*l> Substituting the value for A Nb-95.GI-LLI the equation simplifies to; 2.52E+04
* L:           Ci           <B.3>
* VOL Dmax = --------------
cw where:
* cw B-3 [ Ci <B.4>
Dmax              = maximum organ dose (mrem)
* *
A Nb-95,GI-LLI    = 1.51E+06, Gi-LLI ingestion         dose conversion factor for Nb-95 Cmrem/hr per uCi/*l>
* Salem ODCH Rev. 3 07/30/87 Tritium is not included in the limited analysis dose assessment for liquid releases.
Substituting the value for A Nb-95.GI-LLI the equation simplifies to; 2.52E+04
because the potential dose resulting from normal reactor releases is relatively negligible.
* VOL Dmax   = -------------- *         [  Ci                   <B.4>
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.
cw
This amounts to 0.08X of the design objective dose of 3 mrem/yr. Furthermore.
* B-3
the release of tritium is a function of operating time and power level and is essentially unrelated to radwaste system operation
 
* B-4 Sale1 ODC" Rev. 3 07/30/87 Table B-1 Adult Dose Contributions
Salem ODCH Rev. 3   07/30/87
* Fish and Invertebrate Pathways Unit 1 1986 1985 1984 -----------------------------------
* Tritium releases.
-----------------------------------
is   not   included in the limited analysis dose   assessment   for because the potential dose resulting from normal reactor releases relatively negligible.
------------------------------------
is approximately 350 curies.
Radio-Release TB 6I-LLI -Liver Re lease TB 6I-LLI Liver Release TB 6I-LLI Liver nuclide !Cil Dose Dose Dose (Cil Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac.
liquid The average annual tritium release from each Salem Unit The calculated total body dose from such a release is is  2.4E-03   mrem/yr   via the fish and invertebrate ingestion   pathways. This amounts to 0.08X 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 *
Frac. Frac. Frac. Frac. Frac. Frac. ------------------------Fe-59 2.40E-03 0.01 0.02 0.03 1.40E-03 0.01 0.04 0.04 5.83E-03 0.05 0.04 0.18 Co-58 2.22 0.25 0.42 0.10 6.60E-01 0.12 0.36 0.05 1.58 0.25 0.21 0.02 Co-60 3.10E-01 0.10 0.07 0.04 6.50E-01 0.34 0.42 0.15 1.20 0.54 0.42 0.34 Ag-110111 N/D f f f N/D f f f N/D f f f "n-54 1. 90E-01 0.02 0.06 0.10 8.70E-02 0.02 0.08 0.08 1. 93E-01 . 0.03 0.05 0.22 Nb-95 1.80E-02 f 0.42 f 1.30E-03 f 0.09 f 1. 74E-02 f 0.28 f Cs-137 3.60E-01 0.14 0.01 0.32 2. iOE-01 0.22 f 0.33 5.84E-02 0.05 f 0.11 Cs-134 3.40E-01 0.38 f 0.41 1.60E-01 . 0.29 f 0.35 5.06E-02 0.08 f 0.13 Cr-51 6.00E-02 f f f 3.60E-02 f f f 5.30E-02 f f f Total 3.50E+OO 1.80E+OO 3.33E+OO --------------------------
* B-4
f less than 0.01 N/D = not detected
 
* Table B-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1986 1985 1984 -----------------------------------
Sale1 ODC" Rev. 3 07/30/87 Table B-1 Adult Dose Contributions Fish and Invertebrate Pathways Unit 1 1986                                     1985                                     1984 Radio-       Release      TB    6I-LLI - Liver    Re lease      TB      6I-LLI Liver      Release      TB      6I-LLI Liver nuclide        !Cil      Dose      Dose    Dose      (Cil      Dose        Dose Dose          (Ci)      Dose      Dose  Dose Frac. Frac. Frac.                      Frac. Frac. Frac.                        Frac. Frac. Frac.
-----------------------------------
Fe-59        2.40E-03    0.01    0.02    0.03      1.40E-03    0.01      0.04  0.04        5.83E-03 0.05          0.04  0.18 Co-58        2.22        0.25    0.42    0.10      6.60E-01    0.12      0.36  0.05        1.58        0.25      0.21  0.02 Co-60        3.10E-01    0.10      0.07    0.04      6.50E-01    0.34      0.42  0.15        1.20        0.54      0.42  0.34 Ag-110111    N/D          f          f      f      N/D            f          f    f        N/D          f          f      f "n-54        1. 90E-01  0.02      0.06    0.10    8.70E-02      0.02      0.08  0.08        1. 93E-01 . 0.03      0.05  0.22 Nb-95        1.80E-02      f      0.42      f      1.30E-03        f        0.09    f        1. 74E-02    f        0.28    f Cs-137      3.60E-01    0.14      0.01    0.32    2. iOE-01    0.22        f  0.33        5.84E-02 0.05            f    0.11 Cs-134        3.40E-01    0.38        f    0.41    1.60E-01 . 0.29        f  0.35        5.06E-02 0.08            f    0.13 Cr-51        6.00E-02      f        f      f      3.60E-02        f          f    f        5.30E-02      f          f      f Total        3.50E+OO                              1.80E+OO                                  3.33E+OO f less than 0.01 N/D = not detected
------------------------------------
* 1986 Table B-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1985                                    1984 Radio-       Release       TB     61-LLI Liver     Release        TB       6I-LLI Liver       Release     TB       61-LLI Liver nuclide         {Cil       Dose     Dose   Dose       !Cil       Dose       Dose Dose         (Cil      Dose       Dose   Dose Frac. Frac. Frac.                       Frac. Frac. Frac.                      . Frac. Frac. Frac
Radio-Release TB 61-LLI Liver Release TB 6I-LLI Liver Release TB 61-LLI Liver nuclide {Cil Dose Dose Dose !Cil Dose Dose Dose (Cil Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. . Frac. Frac. Frac * ------------------------Fe-59 4.00E-03 0.02 0.03 0.05 1.10E-03 0.01 0.02 0.02 7.56E-03 0.08 0.06 0.24 Co-58 3.32 0.32 0.44 0.13 8.40E-01 0.14 0.23 0.06 1.30 0.25 0.21 0.13 Co-60 3.80E-01 0.10 0.06 0.04 6.30E-01 O.JO 0.21 0.13 9.79E-01 0.53 0.41 0.28 Ag-1101 N/D f f f N/D f f f N/D f f f "n-54 2.20E-01 0.02 0.05 0.10 1.10E-01 0.02 0.05 0.09 1.61E-01 O.OJ 0.05 0.18 Nb-95 2.50E-02 f 0.41 f 1.40E-02 f 0.48 f 1.36E-02 f 0.27 f Cs-137 3.70E-01 0.20 f 0.29 2.30E-01 0.2.J f 0.33 4.81E-02 0.05 f 0.10 Cs-134 3.60E-01 0.34 f 0.38 1.80E-01 0.30 f 0.35 2.63E-02 0.05 f 0.07 Cr-51 9.50E-02 f f f 3.50E-02 f f f 3.64E-02 f f f Total 4.77E+OO 2.04E+OO 2.75E+OO --------------------------
* Fe-59        4.00E-03    0.02      0.03    0.05    1.10E-03      0.01       0.02 0.02        7.56E-03   0.08      0.06  0.24 Co-58         3.32        0.32      0.44    0.13    8.40E-01     0.14      0.23  0.06      1.30        0.25       0.21   0.13 Co-60         3.80E-01     0.10     0.06    0.04     6.30E-01     O.JO      0.21  0.13        9.79E-01    0.53      0.41  0.28 Ag-1101      N/D           f         f       f       N/D           f         f     f         N/D           f         f     f "n-54         2.20E-01     0.02     0.05    0.10     1.10E-01      0.02       0.05  0.09        1.61E-01   O.OJ      0.05   0.18 Nb-95         2.50E-02     f       0.41    f       1.40E-02      f         0.48  f         1.36E-02     f         0.27    f Cs-137       3.70E-01     0.20      f      0.29    2.30E-01     0.2.J      f   0.33       4.81E-02     0.05       f     0.10 Cs-134       3.60E-01     0.34      f     0.38    1.80E-01     0.30        f   0.35       2.63E-02     0.05        f     0.07 Cr-51         9.50E-02     f         f       f       3.50E-02       f         f     f         3.64E-02     f         f     f Total         4.77E+OO                               2.04E+OO                                 2.75E+OO f    Jess* than 0.01 N/D = not detected B-5
f Jess* than 0.01 N/D = not detected
 
* B-5 
Salem ODCH Rev. 3 07/30/87 APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent
* *
* C-1
* Salem ODCH Rev. 3 07/30/87 APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent C-1 
 
* * '.
Salem ODCH Rev. 3 07/30/87 APPENDIX C Technical Bases for Effective Dose Factors -
* Salem ODCH Rev. 3 07/30/87 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.
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 eHective 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 CK        H    or N      times the total eU      eH      eH 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 *
These eHective 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 CK H or N times the total eU eH eH 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
* Effective  dose    transfer factors are calculated by the following equations:
* Effective dose transfer factors are calculated by the following equations:
KeH  = L  <Ki
where: KeH Ki fi where: KeH = L <Ki
* fi >                          (c.1) where:
* fi > ( 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 relative to the total noble gas activity CL+ 1.1 H>eH = ["((Li + 1.1 Mil
KeH        =  the effective total body dose factor due to gamma emissions from all noble gases released Ki        =  the total body dose factor due to gamma emissions from each noble gas radionuclide i released fi        =  the fractional abundance of noble gas radionuclide          relative to the total noble gas activity
* fil <C.21 <L + 1.1 HleH = the effective skin dose factor due to beta and gamma emissions from all no_b le gases rel eased <Li + 1.1 Hi> = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released C-2 
'.                         CL+ 1.1 H>eH    = ["((Li + 1.1 Mil
* * *-where: tfeff Hi where: Neff Salem ODCH Rev. 3 07/30/87 Heff = L CHi
* fil              <C.21 where:
* fi > CC.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 Neff = l:<Ni
        <L + 1.1 HleH        =  the effective skin dose factor due to beta and gamma emissions from all no_b le gases rel eased
* fi) CC.4> = the effective air dose factor due to beta emissions all noble gases released Ni = 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.
        <Li + 1.1 Hi>        =  the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released
For the past years. the total noble gas have been limited to 1400 Ci for 19821 900 Ci for 19831 21000 Ci for 19841 21800 Ci for 1985. and 21700 for 1986. Therefore.
* C-2
in order to provide a reasonable basis for the derivation of the effective noble gas dose factors1 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 Tab le C-1. To provide an additional degree 4f conservatism.
Salem ODCH Rev. 3 07/30/87 Heff  =  L  CHi
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.
* fi >                        CC.3)
C-3 
* where:
* *
tfeff Hi
* Salem ODCH Rev. 3 07/30/87 For evaluating compliance with the dose limits of Technical Specification 3.11.2.2, the following simplified equations may be used: D and D where: 3.17E-08 = --------* X/Q
                      =
* Heff a.so 3.17E-08
                      =
* I Qi = --------* X/Q
the effective air dose factor due to gamma emissions from noble gases released the air dose factor due to gamma emissions from each noble radionuclide i released all gas Neff  =  l:<Ni
* Neff * "'[_ Qi a.so <C.S> <C.6> D = air dose due to gamma emissions for the cumulative release of all noble gases Cmrad) D = air dose due to beta emissions for the release of all noble gases (mrad) X/Q = atmospheric dispersion to the controlling site boundary He ff = Neff = Qi = 3.17E-08 = a.so = (sec/m3> S.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.
* fi)                          CC.4>
the dose calculational equations simplify to: D = 3.5E-05
where:
Neff        =  the effective air dose factor due to beta emissions f~om all noble gases released Ni          =  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    rele~ses  have been limited to 1400  Ci  for
* 19821    900 Ci for 19831 21000 Ci for 19841 21800 Ci for 1985. and 21700 for 1986.
Therefore.     in  order  to    provide a reasonable basis for the derivation effective noble gas dose factors1 the primary coolant source term from ANSI N237-of  the 1976/ANS-18.1.       "Source    Term  Specifications*" has been used as    representing    a typical    distribution.       The effective dose factors as derived are      presented  in Tab le C-1.
To provide an additional degree 4f 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-3
 
Salem ODCH Rev. 3 07/30/87 For  evaluating  compliance    with the dose    limits    of    Technical  Specification
* 3.11.2.2, the following simplified equations may be used:
D 3.17E-08
                            = --------
* X/Q
* Heff
* I  Qi              <C.S>
a.so and 3.17E-08 D    = --------
* X/Q
* Neff    * "'[_ Qi              <C.6>
a.so where:
D        =  air dose due to gamma emissions for the cumulative release of all noble gases Cmrad)
D        =  air dose due to beta emissions for the ~umulative release of all noble gases (mrad)
X/Q      = atmospheric dispersion to the controlling site boundary (sec/m3>
He ff    = S.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>
a.so    =  conservatism factor to account for the variability in the effluent data Combining the constants. the dose calculational equations simplify to:
D     = 3.5E-05
* X/Q
* X/Q
* l:&deg; Qi <C.7> and D = 7.0E-05
* l:&deg; Qi                     <C.7>
* X/Q * )__ Qi <C.8> The effective dose factors are used a very limited basis for the purpose of facilitating the timely assessment of radioactive effluent releases.
and D     = 7.0E-05
particularly during periods of computer malfunction where a detailed dose assessment may be unavailable
* X/Q   * )__   Qi                     <C.8>
* C-4
The   effective dose factors are used   o~  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 *
* Noble Gases -Total Body and Skin Radionuclide Kr-85 Kr-88 Xe-133m Xe-133 Xe-135 Total Noble Gases -Air Radionuclide Kr-85 Kr-88 Xe-133m Xe-133 Xe-135 Total 0.01 0.01 0.01 0.95 0.02 0.01 0.01 0.01 0.95 0.02 Salem ODCH Rev. 3 07/30/87 Table C-1 Effective Dose Factors Total Body Effective Dose Factor KeH (mrem/yr per uCi/m3) 1.5E+02 2.5E+OO 3.0E+02 3.6E+01 4.8E+02 Gamma Air Effective Dose Factor HeH (mrad/yr per uCi/m3) 1.5E+02 3.3E+OO 3.4E+02 3.8E+01 5.3E+02 Skin Effective Dose Factor <L+ 1.1 H>eH (mrem/yr per uCi/m3) 1.4E+01 1.9E+02 1.4E+01 6.6E+02 7.9E+01 9.6E+02 Beta Air Effective Dose Factor NeH (mrad/yr per uCi/m3) 2.0E+01 2.9E+01 1.5E+01 1.0E+03 4.9E+01 1.1E+03
* C-4
* Based on Noble gas distribution from ANSI N237-1976/ANSI-18.1, ''Source Term Specifications." C-5
 
* *
Salem ODCH Rev. 3 07/30/87 Table C-1 Effective Dose Factors
* APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent D-1
* Noble Gases - Total Body and Skin Radionuclide Total Body Effective Dose Factor KeH (mrem/yr per uCi/m3)
* *
Skin Effective Dose Factor
* Salem ODCH Rev. 3 07/30/97 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.
                                                                          <L+ 1.1 H>eH (mrem/yr per uCi/m3)
This was performed to provide a simplified method for determining compliance with Technical Specification 3.11.2.3 For the infant age group, the controlling pathway is the grass-milk-cow (g/m/c) pathway. An infant receives a greater radiation dose from the g/m/c pathway than any other pathway. Of this g/m/c pathway, the maximum exposed organ including the total body, is the thyroid, and the highest dose contributor is radionuclide I-131. The results for this evaluation are presented in Table D-1. For purposes of simplifying the details of the dose calculation process, it is consetvative to identify a controlling, significant and radionuclide and limit the calculation process to the use of the dose factor for the organ and radionuclide.
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                            Dose Factor                      Dose Factor HeH                            NeH (mrad/yr per uCi/m3)             (mrad/yr per uCi/m3)
Multiplication of the total release (i.e. cumulative activity for all radionuclides) by this dose conversion factor provides for a dose method that is simplified while also being conservative.
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
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 <1.675E12 m2 mrem/yr per uCi/sec>.
* Based on Noble gas distribution from ANSI N237-1976/ANSI-18.1,     ''Source Term Specifications."
By this approach, the maximum dose. commitment will be overestimated since I-131 has the highest pathway dose factor of all radionuclides evaluated.
* C-5
For evaluating compliance with the dose limits of Technical Specification D-2
 
* *
APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent
* Salem ODCH Rev. 3 07/30/87 3.11.2.3.
* D-1
the following simplified equation may be used: where: Dmax = 3.17E-8
 
Salem ODCH Rev. 3     07/30/97
* The APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent Releases pathway dose factors for the controlling infant,age group were evaluated           to determine   the controlling pathway,     organ and radionuclide.       This a~alysis    was performed   to   provide a   simplified   method   for   determining   compliance     with Technical   Specification   3.11.2.3     For the infant age     group,   the controlling pathway   is the grass-milk-cow (g/m/c) pathway.       An   infant receives   a   greater radiation   dose   from the g/m/c pathway than any other pathway.         Of this     g/m/c pathway,   the maximum exposed organ including the total body, is the thyroid, and the highest   dose   contributor   is radionuclide   I-131. The   results   for this evaluation are presented in Table D-1.
For purposes of simplifying the details of the dose calculation process,           it   is consetvative   to identify a controlling,     dos~  significant   orga~  and radionuclide and limit the calculation process to the use of the dose       convers~on  factor     for the organ and radionuclide. Multiplication of the total release (i.e. cumulative activity   for   all radionuclides) by this dose conversion factor provides         for   a dose cal~ulation  method that is simplified while also being conservative.
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   <1.675E12 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 D-2
 
Salem ODCH Rev. 3     07/30/87 3.11.2.3. the following simplified equation may be used:
* where:
Dmax W
Dmax
                      =
                      =
                            =   3.17E-8
* W
* W
* RI-131
* RI-131
* t_Qi Dmax = maxi111um organ dose Cmre111>
* t_Qi maxi111um organ dose Cmre111>
* W = atmospheric dispersion ,parameters to the controlling X/Q = D/Q = Qi = 3.17E-8 = RI-131 = = location(s) as identified in Table 3.2-4. atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways Csec/m3> atmosperic deposition for vegetation1 milk nad ground plane exposure pathways Cm-2> cumulative release over the period of interest for radioiodines and particulates conversion factor (yr/sec> 1-131 dose parameter for the thyroid for the identified controlling pathway 1.675E12 Cm2 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 the total thyroid dose. It is recognized that for some particulate radioiodines  
* atmospheric dispersion ,parameters to the controlling location(s) as identified in Table 3.2-4.
<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.
X/Q  =  atmospheric dispersion for inhalation pathway and H-3 dose contribution via other pathways Csec/m3>
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.
D/Q  =  atmosperic deposition for vegetation1 milk nad ground plane exposure pathways Cm-2>
the dose will be evaluated based on the predetermined controlling pathways as identified in Table 2-4
Qi          =  cumulative release over the period of interest                 for radioiodines and particulates 3.17E-8      =  conversion factor (yr/sec>
RI-131      =  1-131 dose parameter for the thyroid for the identified controlling pathway
                      =  1.675E12 Cm2 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 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-3
* D-3
* Iac:gd Qc:gana Total Body Liver Thyroid Kidney Lung GI-LLI *
 
* Salem ODCH Rev. 3 a7/3a/S7 Table D-1 Infant Dose Fraction of TQtal Organ and Body Dose eer1::1wers Gc:aaa=C2w=IH1.k a.a2 a.23 a.59 a.a2 a.a1 a.a2 2&#xa3; D2at b f aibl!ID::t fa:t.bwa::t  
Salem ODCH Rev. 3   a7/3a/S7 Table D-1
&#xa3; Grass-Cow-Hilk a.92 Ground Plane a.as Inhalation
* Infant Dose Contribution~
* D-4 YC:QYD.d f hnt a.15 a.14 a.15 a.15 a.a2 a.15 
Fraction of TQtal Organ and Body Dose eer1::1wers Iac:gd Qc:gana                  Gc:aaa=C2w=IH1.k                 YC:QYD.d f hnt Total Body                            a.a2                           a.15 Liver                                a.23                           a.14 Thyroid                              a.59                           a.15 Kidney                                a.a2                           a.15 Lung                                  a.a1                           a.a2 GI-LLI                                a.a2                            a.15 Ec:a~H2n  2&#xa3; D2at   C2nic:i~Yii2n b   f aibl!ID::t
* *
* fa:t.bwa::t Grass-Cow-Hilk Ground Plane
* Salem ODCH Rev. 3 07/30/87 Appendix E Radiological Environmental Monitoring Program Sample Type. Location and Analysis E-1
                                                        &#xa3; a.92 a.as Inhalation
* PATHWAY STATION CODE I. DIRECT lFl 1G3 # 2S2 2El 2F2 2F5 2F6 3El 3F2 3F3 3Gl # 3Hl # 3H3 # 4D2 5Sl 501 5Fl 6S2 6Fl 7Sl 7F2 9El lOSl 1001 10F2 lOGl # # Control Station M P85 183/15 4-dh
* D-4
* E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION COLLECTION METHOD 5.8 miles N of or1g1n 2 TLD's will be collected 18.5 miles N of origin from each location quarterly 0.4 miles NNE of origin 4.4 miles NNE of origin 8.7 miles NNE of origin 7.4 miles NNE of origin 7.3 miles NNE of origin 4.1 miles NE of origin 5.1 miles NE of origin 8.6 miles NE of origin 16.6 miles NE of origin 32 miles NE of origin 110 miles NE of origin 3.7 miles ENE of origin 1.0 mile E of origin 3.5 miles E of origin 8.0 miles E of origin 0.2 miles ESE of or1g1n 6.4 miles ESE of origin 0.12 miles SE of origin 9.1 miles SE of origin 4.2 s of origin 0.14 miles SSW of origin 3.9 miles SSW of origin 5.8 miles SSW of origin Collected quarterly 11.*6 miles SSW of origin Rev.
 
* Page.I of 7 ANALYSES Gamma dose quarterly Gamma-dose quarterly 3 7/30/87
Salem ODCH Rev. 3 07/30/87 Appendix E Radiological Environmental Monitoring Program Sample Type. Location and Analysis E-1
* PATHWAY I. DIRECT (Con't) STATION CODE 10F2 lOGl # llSl 11E2 . llFl 12El 12Fl 13El 13F.4 13F2 13F3 1401 14F2 15F3 16El 16F2 16Gl #
 
* TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 5.8 miles SSW oi or1g1n 11.6 miles SSW of origin 0.07 miles SW of origin
  *
* 5.0 miles SW of origin 5.2 miles SW of origin 4.4 miles WSW of origin 9.4 miles WSW 6r origin . 4.2 miles NE of origin 9.8 miles W of origin 6.5 miles W of origin 9.3 miles W of origin 3.4 miles WNW of origin 6.6 miles WNW of origin 5.4 miles NW of origin 4.1 miles NNW of origin 8.1 miles NNW-of origin 14.8 miles NNW of origin COLLECTION METHOD Collected quarterly
* T~BLE E-1
* Page 2 of 7 'ANALYSES Gamma dose quarterly
* Page.I of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY    STATION CODE        LOCATION              COLLECTION METHOD                ANALYSES I. DIRECT lFl     5.8 miles N of or1g1n        2 TLD's will be collected        Gamma dose 1G3 #   18.5 miles N of origin      from each location quarterly    quarterly 2S2    0.4 miles NNE of origin 2El    4.4 miles NNE of origin 2F2    8.7 miles NNE of origin 2F5    7.4 miles NNE of origin 2F6    7.3 miles NNE of origin 3El      4.1 miles NE of origin 3F2    5.1 miles NE of origin 3F3    8.6 miles NE of origin 3Gl #    16.6 miles NE of origin 3Hl #  32 miles NE of origin 3H3 #    110 miles NE of origin 4D2      3.7 miles ENE of origin 5Sl      1.0 mile E of origin 501      3.5 miles E of origin 5Fl      8.0 miles E of origin 6S2      0.2 miles ESE of or1g1n 6Fl      6.4 miles ESE of origin 7Sl      0.12 miles SE of origin 7F2      9.1 miles SE of origin 9El      4.2 mile~ s of origin lOSl    0.14 miles SSW of origin 1001    3.9 miles SSW of origin 10F2    5.8 miles SSW of origin     Collected quarterly             Gamma-dose quarterly lOGl #  11.*6 miles SSW of origin
# Control Station (in addition to controls listed, two additional program controls are used for internal studies and are referred to as SITE-CAL and SITE-ZERO).
  #  Control Station M P85 183/15 4-dh Rev. 3   7/30/87
M P85 183/15 4-dh Rev. 3 7/30/87
 
* EXPOSURE PATHWAY II. AIRBORNE (a) P A R T I c L A T E s STATION CODE 2S2 2F2 3H3 # 1001 16El # Control Station M P85 183/15 4-dh
  *
* TABLE E-1 ' ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 0.4 miles NNE of origin 8.7 miles NNE of origin 110 miles NE of origin 3.9 miles SSW of origin 4.1 miles NNW of origin COLLECTION METHOD Continous low volume air sampler. Sample collected every week along with filter change.
* TABLE E-1 Page 2 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY    STATION CODE          LOCATION              COLLECTION METHOD    'ANALYSES I. DIRECT (Con't) 10F2       5.8 miles SSW oi or1g1n    Collected quarterly    Gamma dose quarterly lOGl #     11.6 miles SSW of origin llSl       0.07 miles SW of origin 11E2
* Page 3 of 7 TYPE AND FREQUENCY OF ANALYSES beta analysis on each weekly sample. Gamma spectrometry shall be performed if gross beta exceeds ten times the yearly mean of control station value. Gross beta analysis done > 24hr. after sampling to allow for Radon and Theron daughter decay Gamma isotopic analysis on quarterly composite Rev. 3 7 /30/87
* 5.0 miles SW of origin
* EXPOSURE PATHWAY II. AIRBORNE (Con't) ( b) I 0 D I N E STATION CODE 2S2 2F2 3H3 # 16El 1001 # Control Station M P85 183/15 4-dh ** TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 0.4 miles NNE of origin 8.7 miles NNE of origin 110 miles NE of origin 4.1 miles NNW of origin 3.9 miles SSW of origin COLLECTION METHOD A TEDA impregnated charcoal flow-through cartridge connected to.air particulate air sampler and is collected weekly at filter change.
                  . llFl       5.2 miles SW of origin 12El       4.4 miles WSW of origin 12Fl       9.4 miles WSW 6r origin 13El    . 4.2 miles NE of origin 13F.4      9.8 miles W of origin 13F2      6.5 miles W of origin 13F3      9.3 miles W of origin 1401      3.4 miles WNW of origin 14F2      6.6 miles WNW of origin 15F3      5.4 miles NW of origin 16El      4.1 miles NNW of origin 16F2      8.1 miles NNW-of origin 16Gl #    14.8 miles NNW of origin
* Page 4 of 7 TYPE AND FREQUENCY OF ANALYSES Iodine 131 analyses are performed on each weekly sample. Rev. 3 7/30/87
  #  Control Station (in addition to controls listed, two additional program controls are used for internal studies and are referred to as SITE-CAL and SITE-ZERO).
* PATHWAY III. WATER (a) S u R F A c E ( b) G R 0 u N D , ( c) s E D I* M E N T STATION CODE 7El 12Cl # 16Fl 2S3 5Dl 3El 7El 12Cl # 16Fl # Control Station M P85 183/15 4-d.h !:
M P85 183/15 4-dh Rev. 3  7/30/87
* TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 1 mile W of Mad Horse Creek; 4.5 miles SE of origin West bank opposite Artificial Island; 2.5 miles WSW of origin C&D canal; 6.9 miles NNW of origin Fresh water holding tank; 700 feet NNW of origin Local farm; 3.5 miles E of origin Local farm; 4.1 miles NE of origin 1 miles W of Mad Horse Creek 4.5 miles SE of origin West bank opposite Artificial Island; 2.5 miles WSW of origin C&D Canal; 6.9 miles NNW of origin COLLECTION METHOD Sample to be collected monthly providing winter icing conditions allow sample collection Collected monthly A sediment sample is taken
 
* Page 5 of 7 ANALYSES Gamma isotopic analysis monthly H-3 on quarterly composite Gamma isotopic monthly Tritium analysis monthly Gamma isotopic anaylsis -semi-* annually Rev. 3 7 /30/87
      *
* PATHWAY IV. INGESTION
* TABLE E-1 '
{a) M { b) I L K F I s H STATION CODE 2F7 3Gl # 5F2 13E3 14Fl llAl l2Cl # # Control Station M P85 183/15 4-dh
ODCM - SALEM GENERATING STATION Page 3 of 7 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE                                                                        TYPE AND FREQUENCY PATHWAY    STATION CODE      LOCATION            COLLECTION METHOD            OF ANALYSES II. AIRBORNE (a) P          2S2      0.4 miles NNE of origin    Continous low volume air    ~ross  beta analysis A                                              sampler. Sample collected    on each weekly R                                              every week along with filter sample. Gamma T                                             change.                      spectrometry shall I         2F2      8.7 miles NNE of origin                                be performed if c                                                                           gross beta exceeds L                                                                           ten times the A                                                                           yearly mean of T                                                                           control station E                                                                           value.
* TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION 5.7 miles NNE of origin 16.6 miles NE of origin 7.0 miles E of origin 4.9 miles w of origin 5.5 miles WNW of origin Outfall area; approx. 650 feet SW of origin West bank opposite Artificial Island;* 2.5 miles WSW of origin COLLECTION METHOD Sample of fresh milk is collected for each farm semimonthly when cows are on pasture, monthly at other times. Two batch samples of fish are sealed in plastic bag or jar and frozen semiannually or when in in season
s 3H3 #   110 miles NE of origin                                  Gross beta analysis done > 24hr. after sampling to allow for Radon and Theron daughter decay 1001     3.9 miles SSW of origin 16El    4.1 miles NNW of origin                                Gamma isotopic analysis on quarterly composite
* Page 6 of 7 ANALYSES
    # Control Station M P85 183/15 4-dh Rev. 3  7 /30/87
* Gamma isotopic and I-131 analyses on each sample on collection Gamma isotopic analysis of edible portion on collection Rev. 3 7 /30/'d7
 
* PATHWAY IV. INGESTION (Cont'd) ( c) I N v E R T E B R A T E s STATION CODE llAl 12Cl # # Control Station M P85 183/15 4-dh
      *                                      **TABLE E-1 Page 4 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE                                                                        TYPE AND FREQUENCY PATHWAY    STATION CODE      LOCATION           COLLECTION METHOD              OF ANALYSES II. AIRBORNE (Con't)
* TABLE E-1 ODCM -SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM LOCATION Outfall area; approx. 650 feet SW of origin West bank opposite Artificial Island; 2.5 miles WSW of origin COLLECTION METHOD Two batch samples of crab are sealed in a plastic bag oi jar and frozen semiannually or when in season.
( b) I        2S2      0.4 miles NNE of origin   A TEDA impregnated charcoal    Iodine 131 analyses 0                                            flow-through cartridge i~      are performed on D                                            connected to.air particulate  each weekly sample.
* Page 7 of 7 ANALYSES Gamma Isotopic analysis of edible portion on collection Rev. 3 7/30/87
I                                            air sampler and is collected N                                            weekly at filter change.
* * * .. * *e e' i! .. ..... ***** ... FIGURE E-1 OFFSITE SAMPLING LOCATIONS ARTIFICIAL ISLAND \ , \ . Rev. 3 7/30/87 
E        2F2      8.7 miles NNE of origin 3H3 #    110 miles NE of origin 16El    4.1 miles NNW of origin 1001    3.9 miles SSW of origin
* * \
    # Control Station M P85 183/15 4-dh Rev. 3  7/30/87
* FIGURE E-2 ONSITE SAMPLING ARTIFICIAL LOCATIONS
 
* ISLAND \ Rev. 3 7/30/87}}
      *
* TABLE E-1 Page 5 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY      STATION CODE        LOCATION              COLLECTION METHOD          ANALYSES III. WATER (a) S          7El      1 mile W of Mad Horse        Sample to be collected      Gamma isotopic u                  Creek; 4.5 miles SE of      monthly providing winter    analysis monthly R                  origin                      icing conditions allow F                                              sample collection          H-3 on quarterly A                                                                          composite c        12Cl #  West bank opposite E                  Artificial Island; 2.5 miles WSW of origin 16Fl    C&D canal; 6.9 miles NNW of origin
( b) G        2S3      Fresh water holding tank; R                  700 feet NNW of origin 0        5Dl      Local farm; 3.5 miles E     Collected monthly          Gamma isotopic u                  of origin                                                monthly N        3El      Local farm; 4.1 miles                                    Tritium analysis D                  NE of origin                                             monthly
    , ( c) s        7El      1 miles W of Mad Horse      A sediment sample is        Gamma isotopic E                  Creek 4.5 miles SE of       taken semi~annually        anaylsis - semi-*
D                  origin                                                   annually I*        12Cl #  West bank opposite M                  Artificial Island; 2.5 E                  miles WSW of origin N        16Fl    C&D Canal; 6.9 miles NNW T                  of origin
    #    Control Station M  P85 183/15 4-d.h Rev. 3   7 /30/87
 
TABLE E-1 Page 6 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY    STATION CODE      LOCATION              COLLECTION METHOD            ANALYSES IV. INGESTION
{a) M         2F7      5.7 miles NNE of origin    Sample of fresh milk is
* Gamma isotopic and I                                            collected for each farm      I-131 analyses on L                                            semimonthly when cows        each sample on K                                            are on pasture, monthly      collection at other times.
3Gl #    16.6 miles NE of origin 5F2      7.0 miles E of origin 13E3    4.9 miles w of origin 14Fl    5.5 miles WNW of origin
{ b) F        llAl    Outfall area; approx.       Two batch samples of fish    Gamma isotopic I                650 feet SW of origin       are sealed in plastic bag    analysis of edible s                                            or jar and frozen            portion on H                                            semiannually or when in      collection in season l2Cl #  West bank opposite Artificial Island;*
2.5 miles WSW of origin
    # Control Station M P85 183/15 4-dh Rev. 3  7 /30/'d7
 
    *                                        *TABLE E-1 Page 7 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY    STATION CODE      LOCATION              COLLECTION METHOD            ANALYSES IV. INGESTION (Cont'd)
( c) I      llAl    Outfall area; approx.      Two batch samples of crab    Gamma Isotopic N                650 feet SW of origin      are sealed in a plastic      analysis of v                                            bag oi jar and frozen        edible portion E                                            semiannually or when in      on collection R                                            season.
T      12Cl #  West bank opposite E                Artificial Island; B                2.5 miles WSW of origin R
A T
E s
    # Control Station M P85 183/15 4-dh Rev. 3  7/30/87
 
FIGURE E-1 OFFSITE SAMPLING LOCATIONS ARTIFICIAL ISLAND e'..........***** ,
                                      \
* *e          i!
                                        \.
Rev. 3 7/30/87
 
FIGURE E-2 ONSITE SAMPLING ARTIFICIAL LOCATIONS
* ISLAND
                  \
  \
* Rev. 3 7/30/87}}

Latest revision as of 06:53, 3 February 2020

Rev 3 to Offsite Dose Calculation Manual.
ML18093A338
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Issue date: 07/30/1987
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Text

SALEM NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION MANUAL Revision 3 07/30/87

  • (

I PDR R

8709030520 870828 ADOCK 05000272 PDR

SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL )

Table of Contents Introduction * . . . . . . . . . . .. 1 1.0 Liquid Effluents 1.1 Radiation Honitoring Instrumentation and Controls * * * *

  • 2 1.2 Liquid Effluent Honitor Setpoint Determination * * * * *
  • 3 1.2.1 Liquid Effluent Honitors <Radwaste, Steam Generator Blowdown and Service Water> * * * * * * * * * * *
  • 4 1.2.2 Conservative Default Values ************* 5 1.3 Liquid Effluent Concentration Limits - 10 CFR 20 5 1.4 Liquid Effluent Dose Calculations - 10 CFR 50 * * *
  • 7 1.4.1 Hember of the Public Dose - Liquid Effluents
  • 7 1.4.2 Simplified Liquid Effluent Dose Calculation 8 1.5 Secondary Side Radioactive Liquid Effluents -

Dose Calculations During Primary to Secondary Leakage *

  • 10 1.6 Liquid Effluent Dose Projection * * * * * * * * * * * * *
  • 12 2.0 Gaseous Effluents 2.1 Radiation Honitoring Instrumentation and Controls * * *
  • 13 2.2 Gaseous Effluent Honitor Setpoint Determination * *
  • 15 2.2.1 Containment and Plant Honitor **** * *
  • 15 2.2.2 Conservative Default Values *
  • 16 2.3 Gaseous Effluent Instantaneous Dose Rate Calculations - 10 CFR 20 * * * * * * *
  • 18 2.3.1 Site Boundary Dose Rate - Noble Gases **** 18 2.3.2 Site Boundary Dose Rate - Radioiodine and Particulates
  • 19 2.4 Noble Gas Effluent Dose Calculations - 10 CFR 50 **** 21 2.4.1 UNRESTRICTED AREA Dose - Noble Gases * * * * * * * * *
  • 21 2.4.2 Simplified Dose Calculation for Noble Gases 21 2.5 Radioiodine and Particulate Dose Calculations - 10 CFR 50
  • 23 2.5.1 UNRESTRICTED AREA Dose - Radioiodine and Particulates
  • 23 2.5.2 Simplified Dose Calculation for Radioiodines and Particulates . * * * * * * * * * * * * * * * *
  • 24 2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations * * * * * * * * * * * *
  • 25 2.7 Gaseous Effluent Dose Projection * * * * * * * * * * *
  • 28 3.0 Special Dose Analyses 3.1 Doses Due To Activities Inside the SITE BOUNDARY 29 3.2 Doses to HEHBERS OF THE PUBLIC - 40 CFR 190 *
  • 30 3.2.1 Effluent Dose Calculations ** 30 3.2.2 Direct Exposure Determination * *
  • 31 4.0 Radiological Environ*ental Monitoring Program * * * * *
  • 32 4.1 Sampling Progra* * * * * * * * * *
  • 32 4.2 Interlaboratory Comparison Program . * . . . . . 33

---Salem ODCH Rev. 3 07/30/87 Table of Contents - Continued Tables .

1-1 Para*eters for Liquid Alarm Setpoint Determination - U~it 1 ** 37 1-2 Para*eters for Liquid Alarm Setpoint Determination - Unit 2 ** 38 1-3 Site Related Ingestion Dose Commitment Factors. Aio * * * * *

  • 39 1-4 Bioaccumulation Factors <BFi> ~ * * * * * * * * * * * * * * *
  • 41 2-1 Dose Factors for Nob 1e Gases * * * * * * * * * * * * * * * *
  • 44 2-2 Para*eters for Gaseous Alarm Setpoint Determinations - Unit 1
  • 45 2-3 Parameters for Gaseous Alarm Setpoint Determinations - Unit 2
  • 46 2-4 Controlling Locations. Pathways and Atmospheric Dispersion for Dose Calculations * * * * * * * * * * * * * * * * *
  • 49 2-5 Path1111ay Dose Parameters - Atmospheric Re.leases * * * * ** so A-1 Calculation of Effective HPC - Unit 1 * * * * *
  • A-4 A-2 Calculation of Effective HPC - Unit 2 * * * * * * . * * *
  • A-5 B-1 Adult Dose Contribut.ions Fish and Drinking Water Pathways Unit 1 * * * * . * * * * * * * * * * * * * * * * * *
  • B-5 B~2 Adult Dose Contributions Fish and Drinking Water Pathways Unit 2 * * * * * * * * * * * * * * * * * . * . * *
  • B-5 C-5 Effective Dose Factors * * * * * * * * * * *
  • C-5 Appendic ies Appendix A - Evaluation of Conservative. Default HPC Value for Liquid Effluents * * * * *
  • A-1 Appendix B Technical Basis for Effective Dose Factors -

Liquid Radioactive Effluents * * * * * * * * * * *

  • B-1 Appendix C - Technical Bases for Effective Dose Factors ~

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

Sample Type. Location and Analysis * * * * * *

  • D-1

Salem ODCH Rev. 3 07130187 SALEH NUCLEAR GENERATING STATION OFFSITE DOSE CALCULATION HANUAL

  • The Salem Offsite Dose Calculation Hanual COOCH> describes the methodology parameters used in! 1) the calculation of radioactive liquid and 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 501 Appendix I and 40 CFR 190.

Hore 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 made ODCH training to will be maintained at the station for use as a document of accepted methodologies and calculations.

the ODCH calculation methodologies and reference parameters Chang~s as is guide and will be 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 ODCH Rev. 3 07130187

  • The liquid controlling effluent monitoring instrumentation and controls and monitoring normal radioactive material relea*ses at in Salem for accordance with the Salem Radiological Effluent Technical Specifications are summarized as follows:

1> 81Dcm_ian~_8Y12maii~-I~cminaii2n1 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 A181C1and D provide the alarm and isolation function for the Unit 1 steam generator blowdown lines. 2-R19 A181C and D provide this function for Unit 2.

2) 8lDcm_i2nl~1 - The alarm functions ror the Service Water System are provided by the radiation monitors on the Containment Fan Cooler discharges C1-R 13 A181C1D and E for Unit 1 and 2-R 13 A181and C for Unit 2>.

Releases from the secondary system are routed through the Chemical Waste Basin where the effluent is monitored Cwith an alarm function>

by R37 prior to release to the environ*ent.

Liquid radioactive waste flow diagrams with the applicable, associated radiation

  • onitoring instrumentation and controls are presented ae Figures 1-1 and 1-2 for Units 1 and 2, respectively *
  • 2

Salem ODCH Rev. 3 07/30/87 Per the require*ents of Technical Specification 3.3.3.81 alarm setpoints shall be established for the liquid effluent monitoring instrumentation to ensure that the release concentration limits of Specification 3.11.1.1 are met (i.e ** the concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS shall be limited to the concentrations specified in 10 CFR 20, Appendix B, Table II, Column 21 for radionuclides and 2.0E-04 uCi/al for dissolved or entrained noble gases>. The following equation* *ust be aatisfied to meet the liquid effluent restrictions:

C <F+f) (1.1) c { -------

where:

c = the effluent concentration limit of Technical Specification (3.11.1.1> implementing the 10 CFR 20 HPC for the site. in uCi/ml c = the setpoint1 in uCi/ml1 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 li*its of 10 CFR 20 in the UNRESTRICTED AREA f = the flow rate at the radiation monitor location. in volume per unit ti*e* but in the sa*e units as F, below F = the dilution water flow rate as measured prior to the release point.

in volu*e per unit ti*e

[Note that if no dilution is provided, c{ C. Also. note that when <F> is large co*pared to (f), then <F + f) = F.l

  • 3
1. 2.1 ~i~~id_Ef f l~~ni_M2nii2cm_iBad~aai~L-Si~am_~~n~cai2c_e12~d2wnL-~h~mi,al The setpoints for the liquid effluent monitors
  • at the equations:

Sale* Nuclear Generating Station SP ~

are HPCe

  • SEN
  • CW determined

+ bkg by the following

( 1.2)

RR Ni th:

[. Ci HPCe =------------ ( 1.3)

J where:

L- (

-~;~:--;

c*

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

HPCe = an effective HPC value for the mixture of radionuclides in the effluent strea* CuCi/*1>

Ci = the concentration of radionuclide i in the liquid effluent CuCi/*l>*

  • NOTE The concentration *ix must include the
  • HPCi SEN cw =

=

=

the HPC value corresponding to radionuclide i fro* 10 CFR 201 Appendix 81 Table 11, Colu*n 2 <uCi/*1>

the sensitivity value to which the monitor is (cp* per uCi/*l>

the circulating water flow rate (dilution water flow> at the tiae of release (gal/*in) calibrated RR = the liquid effluent release rate (gal/*in) bkg = the background of the *onitor Ccp*)

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 strea* 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 fro* 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> *

  • 4

Salem ODCH Rev. 3 07/30/87 Procedural restrictions prevent simultaneous releases fro* either a single unit or both units into a single Circulating Water Syste* discharge

  • 1.2.2 Conservative alarm aetpoints may be deter*ined through the use of default para*eters. Tables 1-1 and 1-2 su*marize all current default valuea in use for Salem Unit-1 and Unit-21 respectively.

They are based upon the following:

a> substitution of the effective HPC value with a default value of lE-05 uCi/*1 for radwaste releases (refer to Appendix A for justification>:

b) for additional ~onservatism** substitution of the 1-131 HPC value of 3E-07 uCi/ml for the R19 Stea* Generator blowdown monitors* R13 Service Water monitor and R37 Che*ical Waste Basin *onitor:

c> substitutions of the operational circulating water flow with the lowest flow. in gal/*in: and, d) substitutiona of the effluent release rate with the highest allowed rate. in gal/*in.

With pre-established alarm setpoints. it is possible to control the radwaste release rate <RR> to ensure the inequality of equation <1.2> is *aintained under changing values for HPCe and for differing Circulating Water System dilutions.

Technical Specification 3.11.1.1 li*its 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 201 Appendix B. Table 111 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

  • Use of the effective HPC value as derived in Appendix A may be non-conservative for the R19 Steam Generator blowdown monitors and R37 Chem i cal Waste Basin *onitors where 1-131 transfer during primary to secondary leakage may potentially be *ore controlling *
  • 5

Salem ODCM Rev. 3 07130187 radiation monitor alarm setpoints are established as addressed above to ensure

  • that liquid these concentration limits are not exceeded.

release However.

results in an alarm setpoint being exceeded, in the event an evaluation co*Pliance with the concentration limits of Techn.ical Specification 3.11.1.1 may any of be performed using the following equation:

~ 1 ( 1.4) where:

Ci = actual concentration of radionuclide as measured in the undiluted liquid effluent (uCi/*l>

HPCi = the HPC value corresponding to radionuclide from 10 CFR 20, Appendix e, Table lls Colu*n 2 <uCi/ml>

= 2E-04 uCi/*l for dissolved or entrained noble gases RR = the actual liquid effluent release rate (gal/*in>

CW = the actual circulating water flow rate (dilution water flow) at the ti*e of the release <gal/min)

  • 6

Salem ODCH Rev. 3 07/30/87

  • 1.4.1 3.11.1.2 radioactive limits the materials dose or dose commitment to HEHBERS Technical OF THE in liquid effluents from each unit of the Sale11 Specification PUBLIC fro*

Nuclear Generating Station to:

during any calendar quarter:

~ 1. S 11re11 to total body per unit

~ 5.0 mrem to any organ per unit during any calendar year:

~ 3.0 11re11 to total body per unit

~ 10.0 mrem to any organ per unit.

Per the surveillance require11ents of Technical Specification 4.11.1.2. the following calculation 11ethode 11ay be used for deter11ining the dose or dose co1111it11ent due to the liquid radioactive effluents fro* Sale11.

1.67E-02

  • VOL Do = -------------- *[ .CCi
  • Aio) ( 1.5)

CW where:

Do = dose or dose com11it11ent to organ O* including total body (mrem>

Aio = site-related ingestion dose co*11itment factor to.the total body or any organ o for rad i onuc l i.de i ( 11re11/hr per uC i /111 l >

Ci = average concentration of radionuclide i* in undiluted liquid effluent representative of the volu11e VOL CuCi/111)

VOL = voluae of liquid effluent released (gal>

cw = average circulating water discharge rate during release period Cga1/11in) 1.67E-02 = conversion factor Chr/11in)

The site-related ingestion dose/dose commitment factors CA > are presented in io Table 1-3 and have been derived in accordance with of NUREG-0133 by the equation:

  • 7

Salem UU~M Rev. j U//jU/~/

  • where:

Aio =

Aio = 1.14E+05 CCUI

  • Bfi)l DFi composite dose parameter for the total body or critical organ o of an adult for radionuclide i1 for the fish and invertebrate ingestion pathways <*rem/hr per uCi/*l>

( 1.6) 1.14E+05 = conversion factor CpCi/uCi

  • ml/kg + hr/yr)

UI = adult invertebrate consumption CS kg/yr)

Bii = bioaccu*ulation factor for radionuclide i in invertegrates from Table 1-4 CpCi/kg + pCi/l)

UF = adult fish consu11ption <21 kg/yr)

BFi = bioaccumulation factor for radionuclide in fish from Table 1-4 CpCi/kg per pCi/1)

DFi = dose conversion factor for nuclide i for adults in pre-selected organ, 01 from Table E-11 of Regulatory Guide 1.109 <mrem/pCi)

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

In lieu of the individual radionuclide dose assessment as presented in Section 1.4.1, the following si*plified dose calculational 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.>

1.21E+03

  • VOL Dtb = --------------

cw

  • L_ Ci ( 1. 7)
  • 8

Salem ODCH Rev. 3 07/30/87

  • D*ax 2.52E+D4
  • VOL

= --------------

cw

  • ( 1.8) where:

Ci = average concentration of radionuclide i* in undiluted liquid effluent representative of the volu*e VOL <uCi/*l>

VOL = volu*e of liquid effluent released (gal>

cw = average circulating water discharge rate during release period (gal/*in)

Dtb - conservatively evaluated total body dose <*re*>

D*ax = conservatively evaluated *axi*u* organ dose <*re*>

1.21E+D3 = conversion factor (hr/min) and the conservative total body dose conversion factor*<Fe-59, total body -- 7.27E+04 *rem/hr per uCi/*l >

2.52E+D4 = conversion factor <hr/*in) and the conservative 11axi11tu* organ dose conversion factor <Nb-95, GI-LLI -- 1.51E+06 *re*/hr per uC il*l >

  • 9

Salem OOCM Rev. 3 07/30/87 1.5 S~'2ndac~--Sid~ __ Badi2a,ii~~-Li~~id_E££l~~nia_and_Q2m~-~al,~laii2na_Q~cing ecimac~--i2_S~'2ndac~-L~akag~

  • During leaks>.

periods of primary to secondary leakage (i.e ** steam generator radioactive material will be transmitted from the primary system to the secondary system. The potential exists for the release of radioactive material tube 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 Cheaical 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 radionuclides.

be Based collected on and analyzed for the principal the identified radioactive material levels gamma emitting and the volume of water discharged, the resulting environmental doses may be calculated based on equation <1.5).

Because the release rate from the secondary system is indirect (e.g., SG blowdown is nor*ally routed to condenser where the condensate clean-up syste*

will reaove much of the radioactive *aterial>* samples should be collected from the final releage point (i.e.~ Che*ical Waste Basin) for quantifying the radioactive *aterial 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

  • 10

Salem ODCH Rev. 3 07/30/87 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 *BY be needed to more accurately account for actual releases *

  • 11

Salem OOCH Rev. 3 07/30/87

  • Technical processing Specification syste* be 3.11.1.3 requires that the liquid used to reduce the radioactive material radioactive liquid waste prior to release when the quarterly projected doses exceed:

levels in waste the 0.375 *re* to the total body1 or 1.25 mre* 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 uaed in conjunction or in lieu of the ion exchange syste* for waate processing require-ments in accordance with Technical Specification 3.11.1.3. Theae processing require*ents are applicable to each unit individually. Exceeding the projected dose requiring processing prior to release for one unit does not in itself

  • dictate processing require*ents for the other unit.

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

Dtbp = Otb (91 t d) (1.9)

Omaxp = Dmax <91

  • d> (1.10>

where:

Dtbp = the total body doae projection for current calendar quarter

<*re*>

Dtb = the total body dose to date for current calendar quarter as determined by equation <1.S> or <1.7> <*re*>

Dmaxp = the maximum organ dose projection for current calendar quarter

<*re*>

Dmax = the maximum organ dose to date for current calendar quarter as deter*ined by equation (1.5) or (1.8) <mre*>

d = the number of days to date for current calendar quarter 91 = the number of days in a calendar quarter

  • 12

Salem ODCM Rev. 3 07/30/87

  • /

The gaseous effluent monitoring instrumentation and controls at Salem for controlling and monitoring normal radioactive material releases in accordance with the Radiological Effl~ent Technical Specifications are summarized as follows!

1> Waait_Gaa_H2l~YQ-~~§itm - The vent header gases are collected by the waste*gas holdup system. Gases may ~e 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> C2niDinmtni_fycgt_Dn~_fctaaYctl~D,YYm-Btlitf - 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> flDDi-~tni - The Plant Vent for each respective unit receives discharges from the waste gas hold-up systems 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 2-R16) provide redundant back-up monitor~ng 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>.

  • The R12A monitors also provide ~he safety function of containment isolation in the event of a fuel handling accident during refueling. During HOOE 6 in accordance with Technical Specification 3/4.3.3, Table 3.3-6, the R12A alarm/trip setpoint shall be established at twice backgrounds providing early indication and containment isolation accompanying unexpected increases in containment airborne radioactive material levels indicative* of a fuel degradation. The R41C monitor may al~o provide this function if the. R12A monitor is inoperable during HOOE 6 *
  • 13

Sal~m ODCM Rev. 3 07/30/87 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 *
  • 14

Salem ODCH Rev. 3 07/30/87 2.2.1 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 calculational method. The measured radionuclide concentrations and release rate are used to calculate the fraction of the allowable release rate, as li*ited by Specification 3.11.2.1, by the equation:

FRAC = C4.72E+02

  • X/Q
  • VF* L CCi
  • Ki)J + 500 <2.1>

FRAC = C4.72E+02

  • X/Q *VF* [. CCi *<Li+ 1.1 Hi))J + 3000 <2.2>
  • where:

FRAC = fraction of the allowable release rate based on the identified radionuclide concentrations and the release flow rate X/Q = annual average meteorological dispersion to the controlling site boundary location (sec/m3) ,

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

Ci = concentration of noble gas radionuclide as determined by radioanalysis of grab sa*ple (uCi/c*3>

Ki = total body dose conversion factor for noble gas radionucl i-de

<*re*/yr per uCi/*3, fro* Table 2-1>

Li = beta skin dose conversion factor for noble gas radionuclide (nre*/yr per uCi/*3, from Table 2-1>

Hi = gam*a air dose conversion factor for noble gas radionuclide (mrad/yr per uCi/m3, from Table 2-1) 1.1 = mrem skin dose per mrad gam*a air dose (mrem/mrad) 4.72E+02 = conversion factor <c*3/ft3

  • min/sec>

500 = total body dose rate limit (mre*/yr) 3000 = skin dose rate li*it (*re*/yr)

  • 15

Salem OOCWRev. 3 07/30/87 Based on the more limiting FRAC (i.e ** higher value) as determined above. the alarm setpoints for the applicable monitors <Rl61 R41Cs and/or R12A> may be calculated by the equation:

SP = CAF * '[Ci

  • SEN + FRACJ + bkg (2.3>

where:

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

SEN = monitor sensitivity (cpm per uCi/c*3>

bkg = background of the monitor (Cp*)

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 refulatory 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),

Creek.

will be Any with the remainder 0.5 allocated to increase in AF above 0.5 for the Salem Nuclear Generating coordinated with the Hope Creek Generating Station to ensure that Hope Station the -

combined alloiation factors for all units do not exceed 1.0.

2.2.2 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 val~es presented in Table 2-1 and 2-2 for Units 1 and 21 respectively *

  • 16

Salem ODCH Rev. 3 07/30/87 These values are based upon:

  • the maximu* ventilation Cor purge) flow rate:

a radionuclide distribution* comprised of 95X Xe-133, 2X Xe-135. 11 Xe-133** 11 Kr-88 and 11 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 maxi*um 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 Speci~ications. Table 6
  • 17

- Salem ODCH Rev. 3 07/30/87

  • limits
  • re*/yr1 are the dose rate at the SITE BOUNDARY due to noble gas releaaes total body and ~3000 mrem/yr, skin.

to {500 Radiation monitor alar* aetpoints established to ensure that these release limits are not exceeded. In the event any gaseous releases from the station results in an alarm aetpoint being exceeded. an evaluation of the SITE BOUNDARY doae rate resulting fro* the release may be performed using the following equations:

Otb = X/Q *I:: CKi *Qi) <2.4) and Os = X/Q

  • l:_ CCLi + 1.1Hi)
  • Qi) (2.5) where:

Otb = total body dose rate <*re*/yr)

Da = skin dose rate <*re*/yr)

X/Q = at*ospheric dispersion to the controlling SITE BOUNDARY location c*eec/*3>

Qi = average release rate of radionuclide i over the release period under evaluation CuCi/sec>

Ki = total body dose conversion factor for noble gas radionuclide Cmre*/yr.per uCi/*31 fro* Table 2-1>

Li = beta skin dose conversion factor for noble gas radionuclide Cmre*/yr per uCi/*31 fro* Table 2-1>

Hi = gamma air dose conversion factor for noble gas radionuclide Cmrad/yr per uCi/*31 fro* Table 2-1) 1.1 = 11rem skin dose per 111rad gam*a air dose C111re*/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 3.11.2.1a, following any release exceeding the above prescribed alar* setpointa. Honitor 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 *onitor reading and monitor sensitivity. The 15 *inute averaging is needed to allow for reasonable monitor response to potentially changing radioactive *aterial concentrations and to exclude potential electronic

  • 18-

Salem ODCH Rev. 3 07/30/87 spikes in monitor readings that may be unrelated to radioactive material releases

  • As identified, any electronic spiking monitor responses *BY be
  • excluded from the analysis.

NOTE: For ad*inistrative purposes. more conservative alarm setpoints than those as prescribed 'above may be imposed. However. conditions exceeding these more limiting alar* setpoints do not necessarily indicate radioactive material release rates exceeding the li*its of Technical Specification 3.11.2.1a. 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 de*onstrating compliance with the limits of Specification 3.11.2.1a.

Actual meteorological conditions concurrent with the release period or the default. annual average dispersion parameters as presented in Table 2-4 *BY be used for evaluating the gaseous effluent dose rate.

2.3.2 Technical Specification 3.11.2.1.b limits the dose rate to ~1500 mrem/yr to any organ for 1-131, tritium and particulates with half-lives greater than 8 days. To demonstrate Co*pliance 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 may be used for the dose rate evaluation:

Oo = X/Q

  • l:_ CRi
  • Qi) (2.6) where:

Do = average organ doae rate over the sampling time period (mrem/yr)

X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location for the inhalation. pathway <sec/*3>

Ri = dose parameter for radionuclide i1 (mrem/yr per uCi/*3> for the child inhalation pathway fro* Table 2-5 Qi = average release rate over the appropriate sampling period and analysis frequency for radionuclide i -- I-131, 1-1331 tritiu* or other radionuclide in particulate for* with half-life greater than 8 days <uCi/sec>

  • 19

Salem ODCH Rev. 3 07/30/87 By substituting 1500 mrem/yr for Do and solving for Q, an allowable release rate for 1-131 can be determined

  • Based on the annual average meteorological
  • dispersion <see Table 2-4> and the most limiting potential pathway, and organ (inhalation, child, thyroid -- Ri allowable release rate for 1-131 is 42 uCi/sec.

= 1.62E+07 age 3

group mrem/yr per uCi/m ), the Reducing this release rate by a factor of 4 to account for potential dose contributions from other radioactive particulate material and other release points *ce.g., Hope Creek), the corresponding release rate allocated to each of the Salem units is 10.S 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 1-131 releases in any 7 day period do not exceed 6.3 Ci1 no additional analyses are needed for verifying compliance with the Technical Specification 3.11.2.1.b limits on allowable release rate *

  • 20

Salem ODCH Rev. 3 07130187

~5 gamma-air and ~10 mrad, beta-air and the calendar year limits ~10 mrad, gamma-air and ~20 mrad, beta-air. The limits

~re applicable separately to each unit and a~e not combined site limits. The following equations may be used to calculate the gamma-air and beta-air doses:

Dg = 3.17E-08

  • X/Q *[ <Hi
  • Qi) (2.7) and Db =. 3.17E-08
  • X/Q
  • C <Ni *Qi) (2.8) where:

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

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

X/Q = atmospheric dispersion to the controlling SITE BOUNDARY location

<sec/a3>

=

Qi cumulative release of noble gas radionuclide i over the period of interest CuCi)

Hi = air dose factor due to gamma emissions from noble gas radionuclide i (arad/yr per uCi/m3, from Table 2-1>

Ni = air dose factor due to beta emissions from noble gas radionuclide (mrad/yr per uCi/m3, Table 2-1>

3.17E-08 = conversion factor. (yr/sec>

In lieu of the individual noble gas radionuclide dose assessment as presented above, the following simplified dose calculational equations may 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 aethod.>

  • 21

Salem ODCH Rev. 3 07/30/87 3.17E-08 Og = --------

  • X/Q
  • HeH * 'L Qi (2.9) a.so
  • Ob =

3.17E-08 a.so and

  • X/Q
  • Neff * 't. Qi <2.10) where:

He ff = S.3E+02. effective ga1111a-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 CuCi) a.so = 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 *

  • 22

Salem ODCM Rev. 3 07/30/87 3.17E-08 Dg = --------

  • X/Q
  • Heff
  • I: Qi (2.9) a.so
  • Db = --------

a.so and 3.17E-08

  • X/Q
  • Neff
  • t_ Qi <2.10) where:

He ff = -s.3E+021 effective gamma-air dose factor (mrad/yr per uCi/~ 3 >

Neff = 1.1E+Q3, effective beta-air dose factor (mrad/yr per uCi/m >

Qi = cumulative release for all noble gas radionuclidea CuCi>

o.so = conservatism factor to account for potential variability in the radionuclide distribution Actual meteorological condition* concurrent with the release period or the default, annual average dispersion parameters aa presented in Table 2-4, may be used for the evaluation of the gamma-air and beta-air doaea *

  • 22

Salem ODCH Rev. 3 07/30/87 2.5 Radioiodine and Particulate Dose Calculations - 10 CFR 50

  • 2.5.1 with shall UNRESTRICTED AREA Ooae - Radioiodine and Particulates.

requirements of Technical Specification 3.11.2.31 a In periodic accordance aeaeasment be perforaed to evaluate compliance with the quarterly dose limit of i7.5 mrem and calendar year limit i15 mrem to.any organ. The following equation may be used to evaluate the maximum organ dose due to releases of I-1311 tritium and particulates with half-lives greater than 8 days:

Daop = 3.17E-08

  • W
  • SFp
  • L. CRi
  • Qi> <2.11) where:

Daop = dose or dose commitment via controlling pathway p and age group a

<a* identified in Table 2-4> to organ o. including the total body

<111rem)

= atmospheric dispersion parameter to the controlling location(e) as identified in Table 2-4 X/Q = atmospheric dispersion for inhalation !athway and H-~ dose contribution via other pathways Csec/m )

D/Q = atmospheric deposition fo~ 2 vegetation1 milk and ground plane exposure pathways Cm 3 or (m 2 Ri = dose factor for radionuclide i1 <mrem/yr per uCi/m) mrem/yr per uCi/mec> from Table 2-5 for each age group a and the applicable pathway pas identified in Table 2-4. Values for R.

were derived in accordance with the methods described in NUREG!

0133.

Qi = cumulative releaee over the period of interest for radionuclide i

-- I-131 or radioactive material in particulate form with half-1 ife greater than 8 days CuCi).

SFp = annual *eaeonal correction factor to account for the fraction of the year that the applicable expo*ure pathway does not exist.

1> For *ilk and vegetation exposure pathways:

= A eix month fresh vegetation and grazing season <Hay through October)

= o.s

2) For inhalation and ground plane exposure pathways:

= 1.0 For evaluating the 111axi111um exposed individual 1 the infant age group is controlling for the milk pathway and the child age group is controlling for the

  • 23

Salem ODCH Rev. 3 07/30/87 vegetation pathway. Only the controlling age group and pathway aa identified in

  • Table 2-4 need be evaluated for compliance with Technical Specification 3.11.2.3.

2.5.2 Simplified Dose Calculation for Radioiodinea and Particulatea. In lieu of the individual radionuclide CI-131 and particulates> doae aaseaament aa presented above. the following simplified doae calculational equation may be uaed for -verifying compliance with the doae limita of Technical Specification 3.11.2.3 (refer to Appendix D for the derivation and justification of thia simplified method).

Dmax = 3.17E-08

  • W
  • SFp
  • RI-131 .- [_Qi <2.12) where:

Dmax * = maximum organ dose <*rem>

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

= controlling 1.05E+121 infa~t thyroid doae parameter with the cow-milk pathway (m - mrem/yr per uCi/eec) w = D/Q for radioiodine1 2.lE-10 1/*

Qi = cumulative releaae over the period of interest for radionuclide i -- I-131 or radioactive material in particulate form with half life greater than 8 daya (uCi)

The location of expoeure pathways and the maximum organ do*e calculation may be based on the available pathway* in the surrounding environ*ent of Salem as identified by the annual land-u*e census <Technical Specification 3.12.2>.

Otherwise, the doae will be eva~uated baaed on the 9redetermined controlling pathwaya as identified in Table 2~4 *

  • 24

Salem ODCH Rev. 3 07/30/87 2.6 Secondary Side Radioactive Gaseous Effluents and Dose Calculations

  • During material periods may condensables be of primary to secondary leakage,

<e.g., noble gases) will be minor levels released via the secondary system to the predominately of radioactive atmosphere.

released via Non-the condenser evacuation system and will be monitored and quantified by the routine plant vent Monitoring and sa*pling syetem 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 Ci.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-01331 the releases of the radioiodines and particulates may be calculated by the equation:

Qi = Ci

  • Rsgb
  • Fft
  • C1-SQftv> <Z.13>

where:

Qi = the release rate of radionuclide. i1 from the steaa generator flash tank vent CuCi/aec>

Ci = the concentration of radionuclide. i1 in the secondary coolant water averaged over not *ore than one week CuCi/ml>

Rsgb = the stea* generator blowdown rate to the flash tank (ml/sec>

Fft = the fraction of blowdown flashed in the tank determined fro* 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 0.851 based on NUREG-0017 *

  • 25

Salem ODCH Rev. 3 07/30/87 Tritium releases via the steam flashing may also be quantified using the above

  • equation with the aseumption of a steam quality <SQftv) equal to O. Since the H-3 will be associated with the water molecules. it is not necessary to for the moisture carryover which is the transport media for the radioiodinea and particulates.

account Based on the design and operating conditions at Sale** the fraction of blowdown converted to steam <Fft> is approximately 0.48. The equation simplifies to the following:

Qi = 0.072 Ci Rsgb <2.14)

For H-3, the simplified equation is!

Qi = 0.48 Ci Rsgb (2.15)

Also during reactor shutdown operations with a radioactively contaminated

  • secondary system. radioactive material may be released to the atmosphere via the atmospheric vi~ the reliefs <PORV> and the safety reliefs on the main steam steam driven auxiliary feed pump exhaust. The evaluation lines of and the radioactive material concentration in the steam relative to that in the steam generator water is based on the guidance of NUREG-00171 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 atmo*pheric stea* releases is:
      • 26

Salem ODCH Rev. 3 07/30/87 l

Qi = 0.13

  • L_ CCij
  • SFj) *PF (2.16) where:

Qij =release rate of radionuclide i via pathway j <uCi/sec)

SFj = steam flow for release pathway j

= 4501000 lb/hr per PORV

= 8001000 lb/hr per safety relief valve

= 501000 lb/hr for auxiliary feed pump exhaust PF = partitioning factor, ratio of concentration in steam to that in the water in the steam generator

= 0.01 for radioiodines

= 0.005 for all other particulates

= 1.0 for H-3 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.

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 wouid be to isolate the effected steam generator. thereby reducing the potential releases during the shutdown/cooldown process. Use of the above calculational 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 C2.12) to calculate offsite doses for demonstrating compliance with the Radiological Effluent Technical Specifications *

  • 27

Salem ODCH Rev. 3 07/30/87 2.7 Gaaeoua Effluent Doae Projection

  • Technical SYSTEM material and Specification 3.11.2.4 requires that the GASEOUS RADWASTE VENTILATION EXHAUST TREATMENT SYSTEM be used to reduce levels prior to discharge when projected dosee exceed TREATHENT radioactive one-half the annual deaign objective rate in any calendar quarter, i.e ** exceeding:

0.625 mrad/quarter. gamma air:

1.25 mrad/quarter, beta air: or

~.875 *re*/quarter. maximum organ.

The applicable gaseous proceeeing ayete*a for *aintaining radioactive material releaaee ALARA are the Auxiliary Building normal ventilation eyatem (filtration syete*a M 112 and 3) and the Waate Gaa Decay Tanke ae delineated in Figures 2-3 and 2-4.

Doee projection& are perfor*ed at least once per 31 days by the following equationa:

  • D gp D bp Dmaxp

= 09 * <91+ d)

= Db * (91+ d>

= D*ax * (91+ d>

(2.17>

(2.18)

<2.19) where:

D gp = 9am*a air doee projection for current calendar quarter Cmrad)

Og = 9a**a air doae to date for current calendar quarter aa determined by equation <2.7> or <2.9) <*rad)

D bp = beta air doae projection for current calendar quarter <*rad)

Db = beta air doee to date for current calendar quarter ae determined by equation <2.8) or <2.10) (mrad>

Dmaxp = maximu* organ doae projection for current calendar quarter (mrem>

D*ax = maximu* organ doae to date for current calendar quarter aa determined by equation <2.11) or <2.12> (mre*>

d = nu*ber of daya to date in current calendar quarter 91 = number of daya in a calendar quarter

  • 28

Salem OOCH Rev. 3 07/30/87 3.0 Special Dose Analyses

Re)ease Report <RERR>

the Radioactive submitted within 60 days after January 1 of Effluent each year shall include an aseeesment of radiation doses fro* radioactive liquid and gaseoue effluent* to HEHBERS OF THE PUBLIC due to their activities ineide the SITE BOUNDARY.

There is one location on Artificial Island that is acceesible to HEHBERS OF THE PUBLIC for activi~ies unrelated to PSE&G operational and support activitiea.

This location i* the Second Sun <vieitor*s center> located near the contractors gate for the Salem Nuclear Generating Station.

The calculation Methods ae presented in Section* 2.4 and 2.5 may be used for determining the maxi*um potential dose to a HEHBER OF THE PUBLIC based on the

  • para*eters fro* Table 2-4 and 2 houre per visit per year. The default value for the mereorological dispereion data as presented in Table 2-3 *BY be used if current year *eteorology in unavailable at the ti*e of NRC reportin. However. a follow-up evaluation shall be performed when the data beco*ee available *
  • . 29

Salem ODCH Rev. 3 07/30/87 3.2 Total dose to HEHBERS OF THE PUBLIC - 40 CFR 190 The Radioactive Effluent Releaee Report <RERR> submitted within 60 days after January 1 of each year ahall aleo include an aseess*ent of ihe radiation doae to the likely moat exposed HEHBER OF THE PUBLIC for reactor releaeea and other nearby uranium fuel cycle sources (including doae contributions from effluents and direct radiation from on-site sources>. For the likely most expoeed HEHBER OF THE- PUBLIC in the vicinity of Artificial Ieland1 the source* of exposure need only consider the Sale* Nuclear Generating Station and the Hope Creek Nuclear Generating Station: No other fuel cycle facilities contribute to the HEHBER OF THE PUBLIC doae for the Artificial Ieland vicinity.

The doee contribution from the operation of Hope Creek Nuclear Generating Station will be estimated baaed on the *ethoda as presented in the Hope Creek Offaite Dose Calculation Hanual CHCGS ODCH>

  • As appropriate for de*onstrating/evaluating compliance with the limits of Technical Specification 3.11.4 <40 CFR 190)1 the result* of the environmental monitoring program *ay be used for providing data on actual measured levels of radioactive material in the actual pathways of exposure.

3.2.1 Effluent Doee Calculation*. For purpoees of implementing the surveillance requirements of Technical Specification 3/4.11.4 and the reporting requirements of 6.9.1.11 CRERR>s doae calculations for the Salem Nuclear Generating Station may be performed using the calculational method* contained within thie ODCH: the conservative controlling pathway* and locations of Table 2-4 or the actual pathway* and locations ae identified by the land use ceneue <Technical Specification 3/4.12.2) may be uaed. Average annual meteorological diapereion

  • 30

Salem ODCH Rev. 3 07/30/87 para*eters or *eteorological condition* concurrent with the release period under evaluation may be used *

  • 3.2.2 Direct Exposure Pose Determination. Any potentially significant exposure contribution to off-site individual doees may be evaluated based on the direct results of the environmental *easurementa <e.g.. TLD1 Ion cha*ber meaaurements>

and/or by the uee of a radiation traneport and ehielding calculational method.

Only during atypical condition* will there exiet any potential for significant on-site eources at Sale* that would yield potentially significant off-site doees Ci.e., in excess of 1 mrem per year to a HEHBER OF THE PUBLIC>. that would require detailed evaluation for de*onstrating compliance with 40 CFR 190.

However. should a situation exist whereby the direct exposure contribution is potentially significant. on-site measure*ente. off-site meaeurements and/or calculational techniques will be used for determination of dose for aseesaing 40 CFR 190 compliance *

  • 31

Salem ODCH Rev. 3 07/30/87 4.D Radiological Environ*ental Monitoring Program 4.1 Sa*pling Program

  • The CREHP>

operational phaae of the Radiological Environmental Horiitorin9 Progra11 ia conducted in accordance with the require11enta of Appendix A Technical Spec if icat ion 3 .12. The object i vee of the pro9ra11 are:

- To determine whether *ny ei9nificant increaaea occur in the concentration of radionuclide* in the critical pathway* of exposure in the vicinity of Artificial Ieland:

- To deter11ine i f the operation of the Salem Nuclear Generatin9 Stations ha* reeulted in any increase in the inventory of lon9 lived radionuclidea in the environ~ent:

To detect any changes in the ambient ga1111a radiation levela: and

- To verify that SNGS operation* have no detrimental effecta on the health and eafety of the pub 1 ic or on the environ*ent.

The aa11plin9 require111enta <type of sa*ples, collection frequency and analysis) and sa*ple location* are presented in Appendix E*

  • 32

Salem OOCH Rev. 3 07/30/87 4.2 Interlaboratorv Comparison Proqra*

  • Technical Specification 3.12.3 requires analyses be perforaed material supplied as part of an Interlaboratory Comparison.

on radioactive Participation in an approved Interlaboratory Comparison Prograa provides a check on the precisness 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 *

  • 33

t...... rn'I HU.1,.:100 COlJl.AHI

~ -l(Il SEAL WAIUC FltlER ,

--ft St:A.1 WAIEH

~

ID COMP 10 ACI011A ffAI...

RAIUI ffA1"'9 ffAI-Rw111 RIWlll llU ....,.. ........

~

ro

  • 1r---::--~~H'!.----------r-- ---1=,

L__~

sv

'I I

FOR INFORMATION ONLY If.I iii FUER RADIATION MONITOR

.... nMS1J1 FIGURE 1-1

£::, INSTRUMENT PANEL 17MSl:l1 RADIATION

-#-PNEUMATIC

-ELECTiaC.AL ...... MONITORING UOUID

-+-- UOUID EFFLUENT

-+-GAS EFFLUENT

  • M l

~::*~-;;,~-~*-:___ : : =~~1 lllllAICA~A alC WIR OliDi 121 <XlHlllNSER t=.. -

1----- ~*

CtlNdlHSlff FQ.~~*

~l-L---* fAIWIO FOR INFORMATION ONLY lEOENp ~ *

!El flJ'&I ..... llMSlll 11t£11t1 FIGURE 1-2 Ii RADIATION lllOHITOR 6 INSDIUIENT MHEL ntilS1l1

........

  • RADIATION

_,,_.PNEUllATIC

-EU:CIRICAL ..

'"' *nSUNU

  • AIOI HEADBI ... """"

°"""""'*"'

MONITORING UOUID

-+- IJOUI> EfR.UENT

-+-GAS EfR.UENT ..

111 2UNIT

  • Monitoring for the 124 and 125 containment fan coil units is provided by the Rl3A, B or C monitors through interconnects.
  • WU HUT OTHER UNIT FROM LAUNDRY & CHEM TKS 2UNIT EOUIPORS WHUTS AUX BlllG SU ... TK RCPHEADTKS FLOOR DRS RC LOOP ORS Wl68 Wl172 Cl/CS HUT RELIEFS RX FLANGE LO Wl.177 RHRSUMl'S ACCUMUUUOR ORS EXCESS LET DOWN FROM llAEVN' Wl70 RWPP I Wl222 I RWST I 1WHUT Wl71 WI.Ill:!

RCPS 1 VVMT Wl 8ti 2VVMT

"' Wl71 CllCSHUT Wl.7 RCDT 3SEALLO I

TO&FROM OTHER UNIT Wl..84 2WHUT FC COHTSUMP Wl73 WASTEEVAP

~- ...I FEEOPUMP Wl.178 WASTEEVAP:

Wl.11 I I

Wl.47

'f' 2Wl.174 2Wl.193

~-------~------~---w-+--W-,-"*'--4~

~

WASTE Wl.11 I EVAP I Wl.77 I

I I Wl.1711 I

L----- OllERllOARO ORWMTS Wl.183 2WMHUT Wl.115 TOANDFROM PUMP IWi.2611 1---ciO-+ OTHER UNIT 12 2lWl6 -

FOR INFORMATION ONLY II SW222 ORC WAJER DISCH 21 SERVICE WAJER -~*"--r~~ FIGURE 1-3 LIQUID I.WI&

WASTE ILW41

Sale1 ODc"* Rev. 3 07/30/87 Table 1-1 Para1eters for Liquid Alar1 Setpoint Deter1inations Unit 1 Para1eter Actual Default Units Co11ents Value Value "Pee calculated 1E-05 1. uCi/ml calculate for each batch to be released KPCI-131 3E-07 N/A uCi/11 I-131 "PC conservatively used for SG blow-down and Service Water 1onitor setpoints Ci measured N/A uCi/11 . taken from ga11a spectral analysis of liquid effluent KP Ci as N/A uCi/11 taken fro1 10 CFR 20, Appendix 8, Table II, deter1ined Col. 2.

SEN 1-R18 as 2.9E+07 CPI per uCi/111 radwaste effluent !Cs-137) deter1ined 1-R19 2.9E+07 Stea1 Generator blowdown !Cs-137!

!A,8,C,Dl 1-R13 1.2E+08 Service Water - Contain1ent fan cooling

!A,B,C,D,El  !Cs-137!

cw as 1.85E+05 gp; Circulating Water Syste1, single CW pu1p deter1ined RR 1-R18 as 120 .. gp1 deter11ined prior to release; release rate deter1ined can be adjusted for Technical Specification CDIPliam 1-R19 80 Stea11 Generator blowdown rate per generator 1-R13 2500 Service Water flow rate for Contain1ent fan coolers SP 1-R18 calculated 4.4E+05!+bkgl CPI Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and 1-R19H calculated 2.0E+04!+bkgl desirable for ensuring regulatory co1pliance and for 1aintaining releases ALARA.

1-R13H calculated 2.6E+03!+bkgl Refer tu Appendix Afor derivation ff The KPC values of 1-131 IJE-07 uCi/mll has been used for derivation of the R19 Stea1 Generator blowdown and R13 Service Water monitor setpoints as discussed in Section 1.2.2

  • 37

Sale1 OOC" Rev. 3 07/30/87 Para1eters for Liquid Alarm Setpoint Deter1inations Unit 2

  • Para1eter "PCe "PCI-131 Actual Value calculated 3E-07 Default Value 1E-05 1 N/A Units uCi/11 uCi/11 Co11ents calculate for each batch to be released I-131 "PC conservatively used for SG blow-down, Service Water and Che1ical Waste Basin 1onitor setpoints Ci 1easured N/A uCi/11 taken fro1 ga11a spectral analysis of liquid effluent "PCi as N/A uCi/11 taken fro1 10 CFR 20, Appendix B, Table II, deter1ined Col. 2.

SEN 2-R18 as 8.8E+07 cp1 per uCi/11 radwaste effluent !Cs-137) deter1ined 2-R19 8.8E+07 Stea1 Generator blowdown !Cs-137) fA,B,C,Ol 2-R13 8.8E+07 Service Water - Contain1ent fan cooling fA,B,C,D,El  !Cs-137)

R37 8.8E+07 Che1ical Waste Basin (Cs-137) cw as 1.85E+OS gp1 Circulating Water Syste1, single CW pu1p deter1ined

  • RR 2-R18 2-R19 as deter1ined 120 80 gp1 deter1ined prior to release; release rate can be adjusted for Technical Specification co1pliance Stea1 Generator blowdown rate per generator
  • 2-RtJ 2500 Service Water flow rate for Contain1ent fan coolers R37 300 Che1ical Waste Basin discharge SP 2-R18 calculated. 8.0E+OS!+bkg) cp1 Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and
  • 2-Rt911 calculated 6.1E+04!+bkg) desirable for ensurfng regulatory co1pliance and for 1aintaining releases ALARA.

2-Rt311 calculated 1.9E+OJ!+bkg)

R371111 calculated t.6E+04!+bkgl 1 Refer to Appendix Afor derivation 11 Based on Cs-137 response 111 Actual calculated setpoint for 2-R18 (l.3E+06l is greater than the full scale 1onitor indicator, therefore, for conservatis1 the reco11ended setpoint has been reduced to 8.0E+OS cp1 1111 The "PC value of 1-131 l3E-07 uCi/1ll has been used for derivation of the R19 Stea1 generator .

blowdovn, R13 Service Water and the R37 Che1ical Waste Basin 1onitor setpoints as discussed in

  • Section 1.2.2 38

Salem ODCH Rev. 3 07/30/87 Table 1-3

  • Nuclide Site Related Ingestion Dose Commitment Factors
  • Aio Bone Liver (mrem/hr per uCi/ml)

T.Body Thyroid Kidney H-3 2.82E-1 2.82E-1 2.82E-1 2.82E-1 Lung 2.82E-1 Gl-LLI 2.82E-1 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-1 4.57E-1 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 Hn-54 7.06E+3 1.35E+3 2.10E+3 2 .16E+4 Hn-56 1.78E+2 3 .15E+1 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 1.22E+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+1 1. 20E+1 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+1 4.26E+2 9.85E+1 Br-82 4.07E+O 4.67E+O Br-83 7.25E-2 1.04E-1 Br-84 9.39E-2 7.37E-7 Br-85 3.86E-3 Rb-86 6.24E+2 2.91E+2 1.23E+2 Rb-88 1.79E+O 9.49E-1 2.47E-11 Rb-89 1.19E+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+1 3.71E+O 4.37E+2 Sr-92 3.48E+1 1.51E+O 6.90E+2

.Y-90 6.06E+O 1.63E-1 6.42E+4 Y-91m 5.73E-2 2.22E-3 1.68E-1 Y-91 8.88E+1 2.37E+O 4.89E+4 Y-92 5.32E-1 1.56E-2 9.32E+3 Y-93 1.69E+O 4.66E-2 5.35E+4 Zr-95 1.59E+1 5.11E+O 3.46E+O 8.02E+O 1.62E+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 1.11E+O 3.50E+3 Ho-99 1.28E+2 2.43E+1 2.89E+2 2.96E+2 Tc-99m 1.30E-2 3.66E-2 4.66E-1 5.56E-1 1.79E-2 2.17E+1 Tc-101 l.33E-2 1.92E-2 1.88E-1 3.46E-1 9.81E-3 5.77E-14 39

Sale* OOCH Rev. 3 07/30/87 Table 1-3 <cont'd)

  • Nuclide Ru-103 Site Related Ingestion Dose Commitment Factors. Aio Bone 1.07E+2 Liver Cmrem/hr per uCi/ml)

T.Body 4.60E+1 Thyroid Kidney 4.07E+2 Lung Gl-LLI 1.25E+4 Ru-105 8.89E+O 3.51E+O 1.15E+2 5.44E+3 Ru-106 1.59E+3 2.01E+2 3.06E+3 1.03E+5 Rh-103m Rh-106 Ag-110m 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+1 1.80E-1 1.36E+2 1.95E+3 Te-125m 2.17E+2 7.86E+1 2.91E+1 6.52E+1 8.82E+2 8.66E+2 Te-127m 5.48E+2 1.96E+2 6.68E+1 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+1 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+1 1.92E+O Te-131m 1.40E+2 6.85E+1 5.71E+1 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 1-130 3.96E+1 1.17E+2 4.61E+1 9.91E+3 1.82E+2 1.01E+2 1-131 2.18E+2 3.12E+2 1. 79E+2 1.02E+5 5.35E+2 8.23E+1 1-132 1.06E+1 2.85E+1 9.96E+O 9.96E+2 4.54E+1 5.35E+O 1-133 7.45E+1 1.30E+2 3.95E+1 1.90E+4 2.26E+2 1.16E+2 1-134 5.56E+O 1.51E+1 5.40E+O 2.62E+2 2.40E+1 1.32E-2 1-135 2.32E+1 6.08E+1 2.24E+1 4.01E+3 9.75E+1 6.87E+1 Cs-134 6.84E+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 7.8SE+3 4.07E+3 1.35E+3 2.32E+2 Cs-138 6.07E+O 1.20E+1 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+1 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 1.80E-9 Ba-142 1. 72E+O 1.77E-3 1.-0SE-1 1.50E-3 1.00E-3 2.43E-18 La-140 1.57E+O 7.94E-1 2.10E-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 1.08E+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+1 9.59E+O 4.43E+1 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 40

Salem ODCH Rev. 3 07/30/87

  • Elu~ni Table 1-4 Bioaccumulation Factors <BFi >

(pCi/kg per pCi/liter>*

~!llil!!!liei: E.iab ~!llil!!!liei: ln~ei:iebi:!lie H . 9.0E-01 9.3E-01 c 1.8E+03 1.4E+03 Na 6.7E-02

  • 1.9E-01 p 3.0E+03 3.0E+04 Cr 4.0E+02 2.0E+03 Hn. 5.5E+02 4.0E+02 Fe 3.0E+03 2.0E+04 Co 1.0E+02 1.0E+03 Ni 1.0E+02 2.5E+02 Cu 6.7E+02 1. 7E+03 Zn 2.0E+03 5.0E+04 Br 1.5E-02 3.1E+OO Rb - 8.3E+OO 1.7E+01 Sr 2.0E+OO 2.0E+01 y 2.5E+01 1.0E+03 Zr 2.0E+02 8.0E+01 N_b 3.0E+04 1.0E+02 Ho 1.0E+01 1.0E+01 Tc 1.0E+01 5.0E+01 Ru 3.0E+OO 1.0E+03 Rh 1.0E+01 2.0E+03 Ag 3.3E+03 3.3E+03 Sb 4.0E+01 5.4E+OO Te 1.0E+01 1.0E+02 I 1.0E+01 5.0E+01 Cs 4.0E+01 2.5E+01 Ba 1.0E+01 1.0E+01 La 2.5E+01 1.0E+03 Ce 1.0E+01 6.0E+02 Pr 2.5E+01 1.0E+03 Nd 2.5E+01 1.0E+03 w 3.0E+01 3.0E+01 Np 1.0E+01 1.0E+01
  • 41
  • - ~

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Sale1 ODO! Rev. 3 07/30/87

, Table 2-1 Dose Factors for Noble Gases

  • Radionuclide Total Body Dose Factor Ki (1re1/yr per uCi/13)

Skin Dose Factor Li

!1re1/yr per uCi/;3) 6a11a Air Dose Factor "i

Beta Air Dose Factor Ni (1rad/yr per uCi/13) (1rad/yr per uCi/13l Kr-8311 7.56E-02 1. 93E+01 2.88E+02 Kr-851 1.17E+03 1.46E+03 1.23E+03 1.97E+03 Kr-85 1.61E+01 1.34E+03 1. 72E+01 1.95E+03 Kr-87 5.92E+03 9.73E+03 6.17E+03 1.03E+04 Kr-88 L41E+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-1311 9.15E+01 4.76E+02 1.56E+02 1.11E+03 Xe-1331 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-13511 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

Sale1 ODC" Rev. 3 07/30/87

  • P-aratehr Actual Default Table 2-2 Para1eters for Gaseous Alar1 Setpoint Deter1inations Unit-1 Units Co11ents Value Value 3

X/Q calculated 2.2E-06 s~c/11 USNRC Sale1 Safety Evaluation, Sup. 3 3

VF as 1easured 1.25E+05 ft /min Plant Vent - nor1al operation

!Plant or Ventl fan curves J

VF as aeasured J;SE+04 ft /1in Contain1ent purge (Cont. or Purge} fan curves AF coordinated 0.25 unitless Ad1inistrative allocation factor to with HC6S ensure co1bined releases do not exceed release rate li1it for site.

J Ci 1easured NIA uCi/c11 J

Ki nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Tab I e 2-1 J

Li nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Table 2-1 3

ni nuclide specific N/A 1rad/yr per uCi/1 Values fro1 Table 2-1 J

SEN 1-R41CI as 1.6E+07 cp1 per uCi/c1 Plant Vent deter1ined 1-R16 3.6E+07 Plant Vent (redundant}

1-R12A 2.1E+06 Contain1ent SP 1-R41C calculated 3.3E+04(+bkgl CPI Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and 1-R16 calculated 7.4E+04(+bkg} desirable for ensuring regulatory co1pliance and for maintaining releases ALARA.

1-R12A11 calculated 1.SE+04(+bkg}

1 Based on 1ean for calibration with 1ixture of radionuclides 11 Applicable during "ODES 1 through 5. During "ODE 6 (refueling), 1onitor setpoint shall be reduced to 2X background in accordance*vith Tech Spec Table 3.3-6.

Sale1 ODC" Rev. J 07/J0/87

  • Parameter Actual Default Tab 11! 2-J Para1eters for Gaseous Alar1 Setpoint Deter1inations Unit-2 Units Co11ents Value Value J

I/& calculated 2.2E-06 sec/1 licensing technical specification value J

VF as 1easured 1.25E+05 ft /1in Plant Vent - normal operation (Plant or Vent) fan curves J

VF as 1easured J.5E+04 ft /1in Contain1ent purge (Cont. or Purge! fan curves AF coordinated 0.25 unitless Ad1inistrative allocation factor to with HCGS ensure co1bined releases do not exceed release rate li1it for site.

J Ci 1easured N/A uCi/c1 J

Ki nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Table 2-1 J

Li nuclide specific N/A 1re1/yr per uCi/1 Values fro1 Table 2-1 J

"i nuclide specific N/A 1rad/yr per uCi/1 Values fro1 Table 2-1 J

SEN 2-R41C1 as 1.6E+07 cp1 per uCi/c1 Plant Vent deter1ined 2-R16 .J.5E+07 Plant Vent (redundant!

2-R12A 3.3E+07 Containaent SP 2-R41C calculated J.3E+04(+bkg) CPI Default alar1 setpoints; 1ore conservative values 1ay be used as dee1ed appropriate and 2-R16 calculated 7.2E+04(+bkg) desirable for ensuring regulatory co1pliance and for 1aintaining releases ALARA.

2-R12A11 calculated 2.4E+05(+bkg) 1------------------------

Based on 1ean for calibration with 1ixture of radionuclides 11 Applicable during "ODES 1 through 5. During "ODE 6 (refueling), 1onitor setpoints shall be reduced to 21 background in accordance with Tech Spec Table 3.J-6.

46

Sale1 ODC" Rev. 3 07/30/87

  • Tab le 2-4 Controlling Locations, Pathways and AtlDsplieric Dispersion for Dose Calcuiations 1 1t1ospheric Dispersion Technical ----------------------

Specification Location Path11ay(sJ Control ling X/Q D/Q Age Group (sec/13J (1/12) 3.11.2.1a site boundary noble gases N/A .2.2E-06 N/A

!0.83 tile, NJ direct exposure 3.11.2.1b site boundary inhalation child 2.2E-06 NIA (0.83 1ile, NJ 3.11.2.2 site boundary ga11a-air N/A 2.2E-06 N/A

!0.83 1ile, NJ beta-air 3.11.2.3 residence/dairy 1ilk and infant 5.4E-08 2.1E-10 (4.8 1iles, NNEJ ground plane 6.9.1.10 Second sun direct exposure N/A 8.22E-06 N/A

!0.21 1i le/SEJ and inhalation I The identified controlling locations, pathways and at1Dspheric dispersion are fro1 the Safety Evaluation Report, Supple1ent No. 3 for the Sale* Nuclear Generating Station, Unit 2 !NUREG-0517, Dece1ber 19781 *

  • 47

Salem ODCH Rev. 3 07/30/87 Table 2-5 Pathway Dose Factors - Atmospheric Releases RCio), Inhalation Pathway Dose Factors *- ADULT Cmrem/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+1 2.28E+1 1.44E+4 3.32E+3 1.00E+2 Hn-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 1.88E+5 1.06E+4 Co-57 6.92E+2 3.70E+5 3.14E+4 6.71E+2 Co-58 1.58E+3 9.28E+5 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 1.34E+4 1.45E+4 Zn-65 3.24E+4 1.03E+5 6.90E+4 8.64E+5 5.34E+4 4.66E+4 Rb-86 1.35E+5 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 S.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 1.04E+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-110m 1.08E+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+1 2.48E+6 4.06E+5 1.24E+4 Sb-125 5.34E+4 5.95E+2 5.40E+1 1. 74E+6 1.01E+5 1.26E+4 Te-125m 3.42E+3 1.58E+3 1.05E+3 1.24E+4 3.14E+5 7.06E+4 4.67E+2 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 1.1oE+6 3.83E+5 1.58E+3 1-131 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 1.10E+5 Cs-137 4.78E+5 6.21E+S 2.22E+5 7.52E+4 8.40E+3 4.28E+5 Ba-140 3.90E+4 4.90E+1 1.67E+1 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+5 4.64E+2 Nd-147 5.27E+3 6.10E+3 3.56E+3 2.21E+5 1.73E+5 3.65E+2

.pa

  • 48

Salem ODCH Rev. 3 07/30/87 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+1 3.07E+1 2.10E+4 3.00E+3 1.35E+2 Hn-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 t.98E+4 Ni-63 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 1.65E+7 7.6~E+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 1.61E+7 9.60E+5 1.24E+4 Ag-110m 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+1 3.85E+6 3.98E+5 1.68E+4 Sb-125 7.38E+4 8.08E+2 7.04E+1 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 1.39E+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+S 1.13E+6 3.75E+S 1.46E+5 9.76E+3 5.49E+S Cs-136 S.1SE+4 1.94E+S 1.10E+S 1. 78E+4 1.09E+4 1.37E+S Cs-137 6.70E+S 8.48E+5 3.04E+S 1.21E+5 8.48E+3

  • 3.11E+S Ba-140 5.47E+4 6.70E+1 2.28E+1 2.03E+6 2.29E+S 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+S 6.62E+2 Nd-147 7.86E+3 8.56E+3 5.02E+3 3.72E+5 1.82E+5 5.13E+2
  • 49

Sale* ODCH Rev. 3 07/30/87 Table 2-5 (cont'd)

RC io), Inhalation Pathway Dose Factors - CHILD

  • Nuclide H-3 C-14 Bone 3.59E+4 Liver 1.12E+3 6.73E+3 (mrem/yr per uC i/m3)

Thyroid 1.12E+3 6.73E+3 Kidney 1.12E+3 6.73E+3 Lung 1.12E+3 6.73E+3 GI-LLI 1.12E+3 6.73E+3 T.Body 1.12E+3 6.73E+3 P-32 2.60E+6 1.14E+5 4.22E+4 9.88E+4 Cr-51 8.55E+1 2.43E+1 1. 70E+4 1.08E+3 1.54E+2 Hn-54 4.29E+4 1.00E+4 1.58E+6 2.29E+4 9.51E+3 Fe-55 4.74E+4 2.52E+4 1.11E+5 2.87E+3 7.77E+3 Fe-59 2.07E+4 3.34E+4 1.27E+6. 7.07E+4 1.67E+4 Co-57 9.03E+2 5.07E+5 1.32E+4 1.07E+3 Co-58 1. 77E+3 1.11E+6 3.44E+4 3.16E+3 Co-60 1.31E+4 7.07E+6 9.62E+4 2.26E+4 Ni-63 8.21E+5 4.63E+4 2.75E+5 6.33E+3 2.80E+4 Zn-65 4.26E+4 1.13E+5 7.14E+4 9.95E+5. 1.63E+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 Ru-106 Ag-110m Sb-124 Sb-125 2.79E+3 1.36E+5 1.69E+4 5.74E+4 9.84E+4 6.73E+3 1.14E+4 7.40E+2 7.59E+2 2.33E+3 1.26E+2 9.10E+1 7.03E+3 1.84E+5 2 .12E+4 6.62E+5 1.43E+7 5.48E+6 3.24E+6 2.32E+6 4.48E+4 4.29E+5 1.00E+5 1.64E+5 4.03E+4 1.07E+3 1.69E+4 9.14E+3 2.00E+4 2.07E+4 Te-125m 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 1.82E+5 3.04E+3 I-131 4.81E+4 4.81E+4 1.62E+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+S 9.SSE+4 1.45E+4 4.18E+3 1.16E+S Cs-137 9.07E+S 8.25E+S 2.82E+5 1.04E+S 3.62E+3 1.28E+5 Ba-140 7.40E+4 6.48E+1 2.11E+1 1.74E+6 1.02E+S 4.33E+3 Ce-141 3.92E+4 1.95E+4 8.55E+3 5.44E+5 5.66E+4 2.90E+3 Ce-144 6.77E+6 2.12E+6 1.17E+6 1.20E+7 3.89E+5 3.61E+5 Pr-143 1.85E+4 5.55E+3 3.00E+3 4.33E+5 9.73E+4 9.14E+2 Nd-147 1.08E+4 8.73E+3 4.81E+3 3.28E+5 8.21E+4 6.81E+2
  • 50

Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd>

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

Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 6.,47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 6.47E+2 C-14 2.65E+4 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 5.31E+3 P-32 2.03E+6 1.12E+5 1.61E+4 7.74E+4 Cr-51 5.75E+1 1.32E+1 1.28E+4 3.57E+2 8.95E+1 Hn-54 2.53E+4 4.98E+3 1.00E+6 7.06E+3 4.98E+3 Fe-55 1.97E+4 1.17E+4 8.69E+4 1.09E+3 3.33E+3 Fe-59 1.36E+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 L22E+3 7.77E+5 1.11E+4 1.82E+3 Co-60 8.02E+3 4.51E+6 3.19E+4 1.18E+4 Ni-63 3.39E+5 2.04E+4 2.09E+5 2.42E+3 1.16E+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 1.14E+4 Sr-90 4.09E+7 1.12E+7 1.31E+5. 2.59E+6 Y-91 5.88E+5 2.45E+6 7.03E+4 1.57E+4

  • Zr-95 Nb:..95 Ru-103
  • Ru-106 Ag-110m 1.15E+5 1.57E+4 2.02E+3 8.68E+4 2.79E+4 6.43E+3 9.98E+3 7.22E+3 3.11E+4 4.72E+3 4.24E+3 1.07E+5 1.09E+4 1.75E+6 4.79E+5 5.52E+5 1.16E+7 3.67E+6 2.17E+4 1.27E+4 1.61E+4 1.64E+5 3.30E+4 2.03E+4 3.78E+3 6.79E+2 1.09E+4 5.00E+3 Sb-124 3.79E+4
  • 5.56E+2 1.01E+2 2.65E+6 5.91E+4 1.20E+4 Sb-125 5.17E+4 4.77E+2 6.23E+1 1.64E+6 1.47E+4 1.09E+4 Te-125m 4.76E+3 1.99E-t:3 1.62E+3 4.47E+5 1.29E+4 6.58E+2 Te-127m 1.67E+4 6.90E+3 4.87E+3 3.75E+4 1.31E+6 2.73E+4 2.07E+3 Te-129m 1.41E+4 6.09E+3 S.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  ;....

1.06E+3 1.96E+4 Cs-134 3.96E+S 7.03E+5 1.90E+S 7.97E+4 1.33E+3 7.4SE+4 Cs-136 4.83E+4 1.35E+5 S.64E+4 1.18E+4 1.43E+3 S.29E+4 Cs-137 5.49E+5 6 .12E+S 1. 72E+S 7.13E+4 1.33E+3 4.55E+4 Ba-140 5.60E+4 S.60E+1 1.34E+1 1.60E+6 3.84E+4 2.90E+3 Ce-141 2.77E+4 1.67E+4 S.2SE+3 5-.17E+5 2.16E+4 1.99E+3 Ce-144 3.19E+6 1.21E+6 5.38E+S 9.S4E+6 1.48E+5 1.76E+5 Pr-143 1.40E+4 S.24E+3 1.97E+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

  • 51

Salem OOCH Rev. 3 07/30/87 Table 2-5 <cont'd)

  • RC io), Grass-Cow-Hilk Pathway Dose Factors - ADULT Cmrem/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+10 1.06E+9 1.92E+9 6.60E+8 Cr-51 1. 71E+4 6.30E+3 3.80E+4 7.20E+6 2.86E+4 Hn-54 8.40E+6 2.50E+6 2.57E+7 1.60E+6 Fe-SS 2.S1E+7 1. 73E+7 9.67E+6 9.9SE+6 4.04E+6 Fe-59 2.98E+7 7.00E+7 1.9SE+7 2.33E+8 2.68E+7 Co-S7 1.28E+6 3.2SE+7 2.13E+6 Co-S8 4.72E+6 9.S7E+7 1.06E+7 Co-60 1.64E+7 3.08E+8 3.62E+7 Ni-63 6.73E+9 4.66E+8 9.73E+7 2.26E+8 Zn-65 1.37E+9 4.36E+9 2.92E+9 2.7SE+9 1.97E+9 Rb-86 1.35E+5 1.66E+4 S.90E+4 Sr-89 1.4SE+9 2.33E+8 4.16E+7 Sr-90 4.68E+10 1.3SE+9 1.1SE+10 Y-91 8.60E+3 4.73E+6 2.30E+2 Zr-9S 1.07E+S 3.44E+4 5.24E+4 1. 77E+6 1.SOE+S 2.33E+4 Nb-9S 1.41E+4 7.82E+3 7.74E+3 5.0SE+S 1.04E+S 4.21E+3 R-u-103 1.02E+3 3.89E+3 1.19E+S 4.39E+2 Ru-106 2.04E+4 -3.94E+4 1.32E+6 2.58E+3 Ag-110m S.83E+7 S.39E+7 1.06E+8 2.20E+10 3.20E+7 Sb-124 2.S7E+7 4.86E+S 6.24E+4 2.00E+7 7.31E+8 1.02E+7 Sb-12S 2.04E+7 2.28E+5 2.08E+4 1.58E+7 2.2SE+8 4.86E+6 Te-12S111 1.63E+7 S.90E+6 4.90E+6 6.63E+7 6.SOE+7 2.18E+6 Te-127111 4.S8E+7 1.64E+7 -1.17E+7 1.86E+8 1.54E+8 5.58E+6 Te-129m 6.04E+7 2.2SE+7 2.08E+7 2.S2E+8 3.04E+8 9.57E+6 I-131 2.96E+8 4.24E+8 1.39E+11 7.27E+8 1.12E+8 2.43E+8 Cs-134 _S.6SE+9 1.34E+10 4.3SE+9 1.44E+9 2.3SE+8 1.10E+10 Cs-136 2.61E+8 1.03E+9 5.74E+8 7.87E+7 1.17E+8 7.42E+8 Cs-137 7.38E+9 1.01E+10 3.43E+9 1.14E+9 1.9SE+8 6.61E+9 Ba-140 2.69E+7 3.38E+4 1.1SE+4 1.93E+4 S.54E+7 1. 76E+6 Ce-141 4.84_E+3 3.27E+3 1.S2E+3 1.2SE+7 3. 71E+2 Ce-144 3.S8E+5 1.50E+S 8.87E+4 1.21E+8 1.92E+4 Pr-143 1.S9E+2 6.37E+1 3.68E+1 6.96E+5 7.88E+O Nd-147 9.42E+1 1.09E+2 6.37E+1 5.23E+5 6.52E+O
  • 52

Sale11 ODCH Rev. 3 07/30/87 Table 2-5 <cont'd>

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

  • mrem/yr per uCi/sec> for others Nuclide Bone Liver Thyroid Kidney Lung GI-LLI T.Body H-3 9.94E+2 9.94E+2 9_.94E+2 9.94E+2 9.94E+2 9.94E+2 C-14 6.70E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 1.34E+5 P-32 3.15E+10 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 Hn-54 1.40E+7 4.17E+6 2.87E+7 2.78E+6 Fe-55 4.45E+7 3.16E+7 2.00E+7 1.37E+7 7.36E+6 Fe-59 5.20E+7 1.21E+8 3.82E+7 2.87E+8 4.68E+7 Co-57 2.25E+6 4.19E+7 3.76E+6 Co-58 7.95E+6 1.10E+8 1.83E+7 Co-60 2.78E+7 3.62E+8 6.26E+7 Ni-63 1.18E+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 9.92E+7 9.60E+6 7.22E+5 6.10E+6 Y-91 1.58E+4 6.48E+6 4.24E+2 Zr-95 1.6SE+3 5.22E+2 7.67E+2 1.20E+6 3.59E+2 Nb-95 1.41E+S 7.80E+4 7.57E+4 3.34E+8 4.30E+4 Ru-103 1.81E+3 6.40E+3 1.52E+5 7.75E+2 Ru-106 3.75E+4 7.23E+4 1.80E+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+S 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 S.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 1.07E+10 Cs-136 4.4SE+8 1. 75E+9 9.53E+8 1.SOE+8 1.41E+8 1.18E+9 Cs-137 1.34E+10 1.78E+10 6.06E+9 2.35E+9 2.53E+8 6.20E+9 Ba-140 4.8SE+7 S.9SE+4 2.02E+4 4.00E+4* 7.49E+7 3.13E+6 Ce-141 1.99E+4 1.3SE+4 6.26E+3 3.62E+S 1.ZOE+S 1.53E+3 Ce-144 6.58E+S 2.72E+S 1.63E+5 1.66E+8 3.54E+4 Pr-143 2.92E+2 1.17E+2 6.77E+1 9.61E+S 1.45E+1 Nd-147 1.81E+2 1.97E+2 1.16E+2 7 .11E+S 1.18E+1
  • 53

Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd>

RCio), Grass-Cow-Hilk Pathway Dose Factors - CHILD Cmrem/yr per uCi/m3) for H-3 and C-14 Cm2

  • mrem/yr per uCi/sec> for others Nuclide Bone Liver Thyroid Kidney Lung Gl-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+10 3.64E+9 2.15E+9 3.00E+9 Cr-51 S.66E+4 1.55E+4 1.03E+5 5.41E+6 1.02E+5 Hn-54 2.09E+7 S.87E+6 1.76E+7 5.58E+6 Fe-55 1.12E+8 5.93E+7 3.35E+7 1.10E+7 1.84E+7 Fe-59 1.20E+8 1.95E+8 S.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 1.59E+9 1.07E+8 1.01E+9 Zn-65 4.13E+9 1.10E+10 6.94E+9 1.93E+9 6.85E+9 Rb-86 8.77E+9 5.64E+8 S.39E+9 Sr-89 6.62E+9 2.56E+8 1.89E+8 Sr-90 1.12E+11 1.51E+9 2.83E+10 Y-91 9.14E+5 2.63E+6 1.84E+5 2.44E+4 Zr-95 3.84E+3 8.45E+2 1.21E+3 8.81E+5 7.52E+2 Nb-95 3.18E+5 1.24E+5 1.16E+5 2.29E+8 8.84E+4 Ru-103 4.29E+3 1.08E+4 1.11E+5 1.65E+3 Ru-106 9.24E+4 1.25E+5 1.44E+6 1.15E+4 Ag-110m 2.09E+8 1.41E+8 2.63E+8 1.68E+10 1.13E+8 Sb-124 1.09E+8 1.41E+8 2.40E+5 6.03E+7 6.79E+8 3.81E+7 Sb-125 8.70E+7 1.41E+6 8.06E+4 4.85E+7 2.08E+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 S.93E+8 1.68E+8 2.47E+7 Te-129m 2.72E+8 7.61E+7 8.78E+7 8.00E+8 3.32E+8 4.23E+7 1-131 1.30E+9 1.31E+9 4.34E+11 2.1SE+9 1.17E+8 7.46E+8 Cs-134 2.26E+10 3.71E+10 1.15E+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 1.01E+10 3.62E+9 1.93E+8 4.55E+9 Ba-140 1.17E+8 1.03E+5 3.34E+4 6.12E+4 5.94E+7 6.84E+6 Ce-141 2.19E+4 1.09E+4 4.78E+3 1.36E+7 1.62E+3 Ce-144 1.62E+6 5.09E+5 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+1 Nd-147 4.45E+2 3.60E+2 1.98E+2 5. 71E+5 2.79E+1
  • 54

Salem ODCH Rev. 3 07/30/87 Table 2-5 (cont'd)

RCio), Grass-Cow-Hilk 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+11 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 Hn-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 1.16E+8 1.88E+8 1.55E+8 Co-57 8.95E+6 3.05E+7 1.46E+7 Co-58 2.43E+7 6.05E+7 6.06E+7 Co-60 8.81E+7 2 .10E+8 2.08E+8 Ni-63 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 1.10E+10 Sr-89 1.26E+10 2.59E+8 3.61E+8 Sr-90 1.22E+11 1.52E+9 3.10E+10 Y-91 7.33E+4 5.26E+6 1.95E+3
  • Zr-95 Nb-95 Ru-103 Ru-106 Ag-110m 6.83E+3 5.93E+5 8.69E+3 1.90E+5 3.86E+8 1.66E+3 2.44E+5 2.82E+8 1.79E+3
1. 75E+5 1.81E+4 2.25E+5 4.03E+8 8.28E+5 2.06E+8 1.06E+S 1.44E+6 1.46E+10 1.18E+3
1. 41E+5 2.91E+3 2.38E+4 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 1.99E+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+10 1.75E+10 7 .18E+9 1.85E+8 6.87E+9 Cs-136 1.96E+9 S.77E+9 2.30E+9 4.70E+8 8.76E+7 2 .15E+9 Cs-137 S.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+S 7.41E+1 Nd-147 8.82E+2 9.06E+2 3.49E+2 5.74E+5 S.55E+1
  • 55

Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd)

R< io>, Vegetation Pathway Dose Factors - ADULT

  • Nuclide H-3 Bone

<mrem/yr per uCi/m3) for H-3 and C-14

<m2

  • mrem/yr per uCi/sec> for others Liver 2.26E+3 Thyroid l<i dney 2.26E+3 Lung 2.26E+3 GI-LLI 2.26E+3 T.Body 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 Hn-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 i.67E+8 3.14E+9 3.69E+8 Ni-63 1.04E+10 7.21E+8 1.50E+8 3.49E+8 Zn-65 3.17E+8 1.01E+9 6.75E+8 6.36E+8 4.56E+8 Rb-86 2.19E+8 4.32E+7 1.02E+8 Sr-89 9.96E+9 1.60E+9 2.86E+8 Sr-90 6.05E+11 1.75E+10 1.48E+11 Y-91 5.13E+6 2.82E+9 1.37E+5 Zr-95 1.19E+6 3.81E+5 5.97E+5 1.21E+9 2.58E+5
  • Nb-95 Ru-103 Ru-106 Ag-110m Sb-124 1.42E+5 4.80E+6 1.93E+8 1.06E+7 1.04E+8 7.91E+4 9.76E+6 1.96E+6 2.52E+5 7.81E+4 1.83E+7 3.72E+8 1.92E+7 4.80E+8 5.61E+8 1.25E+10

. 3.98E+9 8.08E+7 2.95E+9 4.25E+4 2.07E+6 2.44E+7 5.80E+6 4.11E+7 Sb-125 1.36E+8 1.52E+6 1.39E+5 1.05E+8 1.50E+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 1-131 8.09E+7 1.16E+8 3.79E+10 1.98E+8 3.05E+7 6.63E+7 Cs-134 4.66E+9 1.11E+10 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 1.29E+8 1.62E+5 5.49E+4 9.25E+4 2.65E+8 8.4.3E+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 1.11E+10 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

  • 56

Salem ODCH Rev. 3 07/30/87 Table 2-5 <cont'd>

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

  • mre111/yr per uC i/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+S 2.91E+S 2.91E+S 2.91E+S 2.91E+S 2.91E+S P-32 1.61E+9 9.96E+7 1.35E+8 6.23E+7 Cr-S1 3.44E+4 1.36E+4 8.8S!:'.+4 1.04E+7 6.20E+4 Hn-S4 4.S2E+8 1.3SE+8 9.27E+s* 8.97E+7 Fe-SS 3.2SE+8 2.31E+8 1.46E+8 9.98E+7 S.38E+7 Fe-S9 1.81E+8 4.22E+8 1.33E+8
  • 9.98E+8 1.63E+8 Co-S7 1. 79E+7 3.34E+8 3.00E+7 Co-S8 4.38E+7 6.04E+8 1.01E+8 Co-60 2.49E+8 3.24E+9 S.60E+8 Ni-63 1.61E+10 1.13E+9 1.81E+8 S.4SE+8 Zn-6S 4.24E+8 1.47E+9 9.41E+8 6.23E+8 6.86E+8 Rb.,.86 . 2. 73E+8 4.0SE+7 1.28E+8 Sr-89 1.S1E+10 1.80E+9 4.33E+8 Sr-90 7 .S1E+11 2.11E+10 1.8SE+11 Y-91 7.87E+6 3.23E+9 2.11E+S Zr-9S 1.74E+6 S.49E+S 8.07E+S 1.27E+9 3.78E+S Nb-9S 1.92E+S 1.06E+S 1.03E+S 4.SSE+8 S.86E+4 Ru-i.03 6.87E+6 2.42E+7 S.74E+8 2.94E+6 Ru-106 3.09E+8 S.97E+8 1.48E+10 3.90E+7 Ag-110m 1.S2E+7 1.44E+7 2.74E+7 4.04E+9 8.74E+6 Sb-124 1.SSE+8 2.8SE+6 3.SlE+S 1~3SE+8 3 .11E+9 6.03E+7 Sb-12S 2~14E+8 2.34E+6 2.04E+S 1.88E+8 1.66E+9 S.OOE+7 Te-12Sm 1.48E+8 S.34E+7 4.14E+7 4.37E+8 1. 98E+7 Te-127m S.S1E+8 1.96E+8 1.31E+8 2.24E+9 1.37E+9 6.S6E+7 Te-129m 3.67E+8 1.36E+8 1.18E+8_ 1.54E+9 1.38E+9 S.81E+7 1-131 7.70E+7 1.08E+8 3.14E+10 1.8SE+8 2.13E+7 S.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 1.4SE+7 1.36E+7 1.13E+8 Cs-137 1.01E+10 1.3SE+10 4.S9E+9 1. 78E+9 1.92E+8 4.69E+9 Ba-140 1.38E+8 1.69E+S 5.7SE+4 1.14E+S 2 .13E+8 8.91E+6 Ce-141 2.82E+S 1.88E+S 8.86E+4 S.38E+8 2.16E+4 Ce-144 S.27E+7 2.18E+7 1.30E+7 1.33E+10 2.83E+6 Pr-143 7.12E+4 2.84E+4 1.6SE+4 - 2.34E+8 3.SSE+3 Nd-147 3.63E+4 3.94E+4 2.32E+4 1.42E+8 2.36E+3
  • 57

Salem ODCH Rev. 3 07130187 Table 2-5 <cont'd)

RC 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 Gl-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 3.56E+6 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 7.01E+5 P-32 3.37E+9 1.58E+8 9.30E+7 1.30E+8 Cr-51 6.54E+4 1.79E+4 1.19E+5 6.25E+6 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 1.98E+8 Co-60 3.78E+8 2.10E+9 1.12E+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 1.24E+12 1.67E+10 3.15E+11 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-110m 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 1.96E+8 2.20E+9 1.23E+8 Sb-125 4.99E+8 3.85E+6 4.62E+5 2.78E+8 1.19E+9 1.05E+8 Te-125m 3.51E+8 9.50E+7 9.84E+7 3.38E+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 1-131 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 1.23E+5 6.14E+4 2.69E+4 7.66E+7 9.12E+3 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
  • 58

Salem OOCH Rev. 3 07/30/87 Table 2-5 (cont'd)

RCio), Ground Plane Pathway Dose Factors Cm2

  • mrem/yr per uCi/sec)

Nuclide Any Organ H-3 C-14 P-32 Cr-51 4.68E+6 Hn-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+S Nb-95 1.36E+8 Ru-103 1.09E+8 Ru-106 4.21E+8 Ag-110m 3.47E+9 Te-125* 1.55E+6 Te-127111 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+10 Ba-140 2.05E+7 Ce-141 1.36E+7 Ce-144 6.95E+7 Pr-143 Nd-147 8.40E+6

  • 59

Salem OOCH Rev. 3 07/30/87

  • APPENDIX A Evaluation of Default HPC Value for Liquid Effluents
  • A-1

Salem ODCH Rev. 3 07130187 Appendix A Evaluation of Default HPC Value for Liquid Effluents In accordance with the requirements of Technical Specification <3.3.3.10) 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 HPC value of 10 CFR 201 Appendix B, Table 111 Column 2. The determination of allowable radionuclide concentration and corresponding alarm setpoint is a function of the individual radionuclide distribution and corresponding HPC 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 HPC value for this distribution.

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

L Ci HPCe = ------------ <A. l >

[ (-~;~d where:

HPCe = an effective HPC .value for a mixture of radionuclide CuCi/ml)

Ci = concentration of radionuclide i in the mixture HPCi = the 10 CFR 20, Appendix 81 Table II, Column 2 HPC value for radionuclide i (uCi/ml>

Based on the above equation and the radionuclide distribution in the effluents for past years from Salemi an effective HPC value can be determine. Results are

  • A-2

Salem ODCH Rev. ! 07/30/87 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 1986, is reasonable radwaste to select an HPCe value of lE-05 uCi/ml as discharges. Using this value to calculate the typical default of R18 it liquid alarm t

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 Cas evaluated for each batch release) that is compatible with plant operations (refer to Tables 1-1 and 1-2) *
  • A-3

5ale1 ODC" Rev. 3 07/30/87 Table A.:.1 Calculation of Effective nPC 5ale1 Unit 1

. Activity Released fCiJ

~--------------------------------------------------------------------

Nuclide "PCt

  • 1981 1982 1983 1984 1985 1986

!uCi/11 l Na-24 3E-05 2.4E-02 1.9E-03 5.3E-03 5.6E-03 6.2E-03 9.2E-04 Cl'-51 2E-03 5.9E-D2 1.4E-01 6.2E-02 5.3E-02 3.6E-02 6.0E-02 l'ln-54 1E-04 7.4E-02 2.1E-01 1.6E-01 1.9E-01 8.7E-02 1.9E-01 Fe-59 5E-05 3.8E-03 8.6E-03 4.2E-02 5.8E-03 1.4E-03 2.4E-03 Co-58 9E-05 4.5E-01 1.7 1.8 1.6 6.6E-01 2.22 Co-60 3E:.os 3.2E-01 . 9.1E-01 7.1E-01 1.2 6.SE-01 3.1E-01 ll'-95 6E-05 1.0E-02 1.1E-02 8.0E-03 1.8E-03 3.2E-03 4.3E-03 Nb-95 1E-04 1.8E-02 4.8E-02 2.2E-02 1. 7E-02 1.3E-03 1.8E-02 Nb-91 9E-04 N/OH 9.SE-03 3.6E-04 2.0E-02 7.2E-03 1.5E-03 Tc-991 3E-03 N/D N/D N/D 1.6E-03 N/D N/D 51'-89 3E-06 N/D N/D 1.2E-03 4.2E-04 1.7E-03 3.5E-07 51'-90 3E-07 N/D N/D N/D 2.2E-05 1.7E-04 3.1E-08 *1 I

"o-99 4E-05 2.5E-04 1.0E-03 1.6E-03 1.9E-03 1.0E-04 N/D I Ag-1101

  • 3E-05 N/D 4.7E-03 N/D N/D N/D N/D 5n-113 8E-05 8.9E-04 2.2E-04 3.8E-04 9.4E-04 N/D 3.5E-04 5b-124 2E-05 6.2E-04 8.0E-03 1.4E-02 1. 7E-02 5.7E-03 8.4E-02 5b:.125 1E-04 3.0E-01 6.8E-03 4.4E-02 4.9E-03 N/D 3.6E-02 I-131 3E-07 1.tE-01 6.5E-02 2.4E-02 4.5E-02 *7.9E-02 1.2E-01 I-133 1E-06 8.8E-02 5.SE-03 3.3E-02 2.SE-02 1.9E-02 1.4E-03 I-135 4E-06 N/D 3.SE-04 1.6E-03 1.2E-03 N/D N/D Cs-134 9E-06 6.0E-02 4.0E-02 1.8E-02 5.1E-02 1.6E-01 3.4E-01
  • Cs-137 2E-05 7.6E-02 5.9E-02 3.0E-02 5.8E-02 2.1E-01 3.6E-01 Ba-140 2E:.05 N/D N/D 1.3E-02 2.1E-03 N/D N/D La-140 2E-05 N/D 7.SE-03 1.3E-02 1.6E-02 1.1E-04 3.SE-04 Total Ci 1.59 3.24 3.00 3.32 1.93 3.75

__ Ci __ 4.86E+05 2.83E+05 1.66E+05 2.46E+05 3.42E+05 4.99E+05 "PCi "PCe !uCi/11 l 3.3E-06 1.14E-05 1.80E-05 1.35E-05 5.63E-06 7.51E-06 f l'IPC value fol' unl'estricted al'ea fl'OI 10 CFR 20, Appendix B, Table II, Colu1n 2.

ff N/D - not detected

  • A-4

Sale1 ODC" Rev. 3 07/30/87

  • Table A-2 Calculation of Effective "PC -

Sale1 Unit 2 Activity Released !Ci) -

  • Nuclide "PCI 1981 1982 1983 1984 1985 1986

!uCi/111)

Na-24 3E-05 2.0E-01 1.2E-03 9.2E-03 4.4E-03 3.5E-03 3.6E-03 Cr-51 2E-OJ 9.5E-02 1.1E-01 4.6E-02 3.6E-02 3.5E-02 9.5E-02 "n-54 1E-04 4.4E-02 2.0E-01 1.4E-01 1.6E-01 1.1E-01 2.2E-01 Fe-59 5E-05 5.8E-03 5.6E-03 J.1E-02 7.6E-03 1.1E-03 4.0E-03 Co-58 9E-05 8.1E-01 1.7 1. 7 1.3 8.4E-01 3.32 Co-60 3E-05 2.6E-01 8.6E-01 5.7E:.01 9.8E-01 6.3E-01 3.8E-01 Zr-95 6E-05 1.0E-02 . 9.7E-03 5.2E-'03 1.2E-03 4.6E-03 1.1E-02 Nb-95 1E-04 1.5E-02 2.3E-02 1.6E-02 1.4E-02 1.4E-02 2.5E-02 Nb-97 9E-04 N/DH 1.1E-02 1.1E-02 2.1E-02 5.7E-03 2.7E-03 Tc-991 3E-03 N/D N/D N/D 1.4E-03 N/D N/D Sr-89 3E-06 N/D N/D 3.2E-04 J.2E-04 1.5E-03 4.1E-07 Sr-90 3E-07 N/D N/D N/D 4.1E-05 1.0E-04 3.2E-08 "o-99 4E-05 7.4E-05 1.7E-04 J.OE-03 t.4E-03 N/D N/D Ag-1101 3E-05 N/D 3.9E-03 N/D N/D N/D N/D Sn-113 8E-05 2.6E-04 1.6E-04 5.9E-04 t.2E-03 N/D 1.1E-03

  • Sb-124 Sb-125*

I-131 I-133 I-135 Cs-134 2E-05 1E-04 JE-07 1E-06 4E-06 9E-06 9.1E-O 8.4E-03 2.6E-02 6.0E-04 N/D 1.8E-02 1.0E-02 1.0E-02 1.3E-01 6.0E-03 N/D 5.1E-02 2.0E-02 9.6E-02 J.6E-02 5.4E-02 1.6E-03 2.0E-02 3.0E-02 3.6E-03 4.2E-02 2.6E-02 4.4E-04 2.bE-02 1.2E-03 N/D 8.4E-02 1.2E-02 N/D 1.8E-01 1.2E-01 5.4E-02 t.2E-01 2.6E-03 N/D 3.bE-01 Cs-137 2E-05 2.9E-02 7.6E-02 J.bE-02 4.8E-02 2.3E-01 3.7E-01 Ba-140 2E-05 N/D N/D 9.8E-03 6.6E-03 N/D N/D La-140 *2E-05 N/D 6.7E-03 8.ff-02 3.0E.:~2 N/D 6.9E-04 Total Ci 1.52 3.21 ' 2.89 2.74 2.15 5.09

-~i- 1.16E+05 5.00E+05 2.26E+05 2.24E+05

  • 3.56E+05 5.20E+05 "PCi "Pee !uCi/11} 1.31E-05 _ 6.42E-06 t.28E-05 1.22E-05 6.04E-06 9.79E-06 1 "PC value for unrestricted area fro1 10 CFR 20, Appendix 8, Table II, Colu1n 2.

11 N/D - not detected

  • A-5

Salem ODCH Rev. 3 07/30/87

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

Salem OOCH Rev. 3 07/30/87 APPENDIX B Technical Basis for Effective Dose Factors -

  • Liquid Effluent Releases The radioactive liquid effluents for the years 19861 1985, 19841 19831 and 1982 were evaluated to determine the dose contribution of the radionuclide distribution. This analysis was performed to evaluate the use of a limited dose analysis for determining environmental doses, providing a simplified method of determining compliance with the dose limits of Technical Specification 3.11.1.2.

For the radionuclide distribution of effluents from Salem. the controlling organ is the GI-LLI. The calculated Gl-LLI dose is predominately a function of the Fe-59, C0-581 C0-60 and Nb-95 releases. The radionuclides1 Co-58 and C0-60 contribute the large majority of the calculated total body dose. The results of the evaluation for 19861 1985, and 1984 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 calculational 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 use the Nb-95 dose conversion factor Cl.51 E+06 mrem/hr per uCi/ml1 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 C7.27 E+04 mrem/hr per uCi/ml1 total body) is the highest among the identified dominant nuclides *

  • B-2

Salem ODCH Rev. 3 07130187 For evaluating compliance with the dose limits of Technical Specification 3.11.1.2, the following simplified equations may be used:

1.67E-02

  • VOL Dtb = --------------
  • A Fe-59,TB * '[, Ci <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)

Ci = total concentration of all radionuclides <uCi/ml>

CLI = average circulating water discharge rate during release period (gal/rain) 1.67E-02 = conversion factor Chr/min)

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

1.21E+03

  • VOL Dtb = -------------- * <B.2>

cw 1.67E-02

  • VOL
  • A Nb-95,GI-LLI Dmax = ---------------------------
  • L: Ci <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 Cmrem/hr per uCi/*l>

Substituting the value for A Nb-95.GI-LLI the equation simplifies to; 2.52E+04

  • VOL Dmax = -------------- * [ Ci <B.4>

cw

  • B-3

Salem ODCH Rev. 3 07/30/87

is not included in the limited analysis dose assessment for because the potential dose resulting from normal reactor releases relatively negligible.

is approximately 350 curies.

liquid The average annual tritium release from each Salem Unit The calculated total body dose from such a release is is 2.4E-03 mrem/yr via the fish and invertebrate ingestion pathways. This amounts to 0.08X 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

Sale1 ODC" Rev. 3 07/30/87 Table B-1 Adult Dose Contributions Fish and Invertebrate Pathways Unit 1 1986 1985 1984 Radio- Release TB 6I-LLI - Liver Re lease TB 6I-LLI Liver Release TB 6I-LLI Liver nuclide !Cil Dose Dose Dose (Cil Dose Dose Dose (Ci) Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac. Frac.

Fe-59 2.40E-03 0.01 0.02 0.03 1.40E-03 0.01 0.04 0.04 5.83E-03 0.05 0.04 0.18 Co-58 2.22 0.25 0.42 0.10 6.60E-01 0.12 0.36 0.05 1.58 0.25 0.21 0.02 Co-60 3.10E-01 0.10 0.07 0.04 6.50E-01 0.34 0.42 0.15 1.20 0.54 0.42 0.34 Ag-110111 N/D f f f N/D f f f N/D f f f "n-54 1. 90E-01 0.02 0.06 0.10 8.70E-02 0.02 0.08 0.08 1. 93E-01 . 0.03 0.05 0.22 Nb-95 1.80E-02 f 0.42 f 1.30E-03 f 0.09 f 1. 74E-02 f 0.28 f Cs-137 3.60E-01 0.14 0.01 0.32 2. iOE-01 0.22 f 0.33 5.84E-02 0.05 f 0.11 Cs-134 3.40E-01 0.38 f 0.41 1.60E-01 . 0.29 f 0.35 5.06E-02 0.08 f 0.13 Cr-51 6.00E-02 f f f 3.60E-02 f f f 5.30E-02 f f f Total 3.50E+OO 1.80E+OO 3.33E+OO f less than 0.01 N/D = not detected

  • 1986 Table B-2 Adult Dose Contributions Fish and Invertebrate Pathways Unit 2 1985 1984 Radio- Release TB 61-LLI Liver Release TB 6I-LLI Liver Release TB 61-LLI Liver nuclide {Cil Dose Dose Dose !Cil Dose Dose Dose (Cil Dose Dose Dose Frac. Frac. Frac. Frac. Frac. Frac. . Frac. Frac. Frac
  • Fe-59 4.00E-03 0.02 0.03 0.05 1.10E-03 0.01 0.02 0.02 7.56E-03 0.08 0.06 0.24 Co-58 3.32 0.32 0.44 0.13 8.40E-01 0.14 0.23 0.06 1.30 0.25 0.21 0.13 Co-60 3.80E-01 0.10 0.06 0.04 6.30E-01 O.JO 0.21 0.13 9.79E-01 0.53 0.41 0.28 Ag-1101 N/D f f f N/D f f f N/D f f f "n-54 2.20E-01 0.02 0.05 0.10 1.10E-01 0.02 0.05 0.09 1.61E-01 O.OJ 0.05 0.18 Nb-95 2.50E-02 f 0.41 f 1.40E-02 f 0.48 f 1.36E-02 f 0.27 f Cs-137 3.70E-01 0.20 f 0.29 2.30E-01 0.2.J f 0.33 4.81E-02 0.05 f 0.10 Cs-134 3.60E-01 0.34 f 0.38 1.80E-01 0.30 f 0.35 2.63E-02 0.05 f 0.07 Cr-51 9.50E-02 f f f 3.50E-02 f f f 3.64E-02 f f f Total 4.77E+OO 2.04E+OO 2.75E+OO f Jess* than 0.01 N/D = not detected B-5

Salem ODCH Rev. 3 07/30/87 APPENDIX C Technical Bases for Effective Dose Factors Gaseous Radioactive Effluent

  • C-1

Salem ODCH Rev. 3 07/30/87 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 eHective 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 CK H or N times the total eU eH eH 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 *

  • Effective dose transfer factors are calculated by the following equations:

KeH = L <Ki

  • fi > (c.1) where:

KeH = the effective total body dose factor due to gamma emissions from all noble gases released Ki = the total body dose factor due to gamma emissions from each noble gas radionuclide i released fi = the fractional abundance of noble gas radionuclide relative to the total noble gas activity

'. CL+ 1.1 H>eH = ["((Li + 1.1 Mil

  • fil <C.21 where:

<L + 1.1 HleH = the effective skin dose factor due to beta and gamma emissions from all no_b le gases rel eased

  • = the skin dose factor due to beta and gamma emissions from each noble gas radionuclide i released
    • C-2
    Salem ODCH Rev. 3 07/30/87 Heff = L CHi
    • fi > CC.3)
    • where:
    tfeff Hi = = the effective air dose factor due to gamma emissions from noble gases released the air dose factor due to gamma emissions from each noble radionuclide i released all gas Neff = l:<Ni where: Neff = the effective air dose factor due to beta emissions f~om all noble gases released Ni = 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 rele~ses have been limited to 1400 Ci for
    • 19821 900 Ci for 19831 21000 Ci for 19841 21800 Ci for 1985. and 21700 for 1986.
    Therefore. in order to provide a reasonable basis for the derivation effective noble gas dose factors1 the primary coolant source term from ANSI N237-of the 1976/ANS-18.1. "Source Term Specifications*" has been used as representing a typical distribution. The effective dose factors as derived are presented in Tab le C-1. To provide an additional degree 4f 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-3 Salem ODCH Rev. 3 07/30/87 For evaluating compliance with the dose limits of Technical Specification
    • 3.11.2.2, the following simplified equations may be used:
    D 3.17E-08 = --------
    • X/Q
    • Heff
    • I Qi <C.S>
    a.so and 3.17E-08 D = --------
    • X/Q
    • Neff * "'[_ Qi <C.6>
    a.so where: D = air dose due to gamma emissions for the cumulative release of all noble gases Cmrad) D = air dose due to beta emissions for the ~umulative release of all noble gases (mrad) X/Q = atmospheric dispersion to the controlling site boundary (sec/m3> He ff = S.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> a.so = conservatism factor to account for the variability in the effluent data Combining the constants. the dose calculational equations simplify to: D = 3.5E-05
    • X/Q
    • l:° Qi <C.7>
    and D = 7.0E-05
    • X/Q * )__ Qi <C.8>
    The effective dose factors are used o~ 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-4
    Salem ODCH Rev. 3 07/30/87 Table C-1 Effective Dose Factors
    • Noble Gases - Total Body and Skin Radionuclide Total Body Effective Dose Factor KeH (mrem/yr per uCi/m3)
    Skin Effective Dose Factor <L+ 1.1 H>eH (mrem/yr per uCi/m3) 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 Dose Factor Dose Factor HeH NeH (mrad/yr per uCi/m3) (mrad/yr per uCi/m3) 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-5
    APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent
    • D-1
    Salem ODCH Rev. 3 07/30/97
    • The APPENDIX D Technical Basis for Effective Dose Parameter Gaseous Radioactive Effluent Releases pathway dose factors for the controlling infant,age group were evaluated to determine the controlling pathway, organ and radionuclide. This a~alysis was performed to provide a simplified method for determining compliance with Technical Specification 3.11.2.3 For the infant age group, the controlling pathway is the grass-milk-cow (g/m/c) pathway. An infant receives a greater radiation dose from the g/m/c pathway than any other pathway. Of this g/m/c pathway, the maximum exposed organ including the total body, is the thyroid, and the highest dose contributor is radionuclide I-131. The results for this evaluation are presented in Table D-1.
    For purposes of simplifying the details of the dose calculation process, it is consetvative to identify a controlling, dos~ significant orga~ and radionuclide and limit the calculation process to the use of the dose convers~on factor for the organ and radionuclide. Multiplication of the total release (i.e. cumulative activity for all radionuclides) by this dose conversion factor provides for a dose cal~ulation method that is simplified while also being conservative. 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 <1.675E12 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 D-2 Salem ODCH Rev. 3 07/30/87 3.11.2.3. the following simplified equation may be used:
    • where:
    Dmax W Dmax = = = 3.17E-8
    • W
    • RI-131
    • t_Qi maxi111um organ dose Cmre111>
    • 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 Csec/m3> D/Q = atmosperic deposition for vegetation1 milk nad ground plane exposure pathways Cm-2> Qi = cumulative release over the period of interest for radioiodines and particulates 3.17E-8 = conversion factor (yr/sec> RI-131 = 1-131 dose parameter for the thyroid for the identified controlling pathway = 1.675E12 Cm2 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 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-3
    Salem ODCH Rev. 3 a7/3a/S7 Table D-1
    • Infant Dose Contribution~
    Fraction of TQtal Organ and Body Dose eer1::1wers Iac:gd Qc:gana Gc:aaa=C2w=IH1.k YC:QYD.d f hnt Total Body a.a2 a.15 Liver a.23 a.14 Thyroid a.59 a.15 Kidney a.a2 a.15 Lung a.a1 a.a2 GI-LLI a.a2 a.15 Ec:a~H2n 2£ D2at C2nic:i~Yii2n b f aibl!ID::t
    • fa:t.bwa::t Grass-Cow-Hilk Ground Plane
    £ a.92 a.as Inhalation
    • D-4
    Salem ODCH Rev. 3 07/30/87 Appendix E Radiological Environmental Monitoring Program Sample Type. Location and Analysis E-1
    • T~BLE E-1
    • Page.I of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY STATION CODE LOCATION COLLECTION METHOD ANALYSES I. DIRECT lFl 5.8 miles N of or1g1n 2 TLD's will be collected Gamma dose 1G3 # 18.5 miles N of origin from each location quarterly quarterly 2S2 0.4 miles NNE of origin 2El 4.4 miles NNE of origin 2F2 8.7 miles NNE of origin 2F5 7.4 miles NNE of origin 2F6 7.3 miles NNE of origin 3El 4.1 miles NE of origin 3F2 5.1 miles NE of origin 3F3 8.6 miles NE of origin 3Gl # 16.6 miles NE of origin 3Hl # 32 miles NE of origin 3H3 # 110 miles NE of origin 4D2 3.7 miles ENE of origin 5Sl 1.0 mile E of origin 501 3.5 miles E of origin 5Fl 8.0 miles E of origin 6S2 0.2 miles ESE of or1g1n 6Fl 6.4 miles ESE of origin 7Sl 0.12 miles SE of origin 7F2 9.1 miles SE of origin 9El 4.2 mile~ s of origin lOSl 0.14 miles SSW of origin 1001 3.9 miles SSW of origin 10F2 5.8 miles SSW of origin Collected quarterly Gamma-dose quarterly lOGl # 11.*6 miles SSW of origin
    1. Control Station M P85 183/15 4-dh Rev. 3 7/30/87
    • TABLE E-1 Page 2 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY STATION CODE LOCATION COLLECTION METHOD 'ANALYSES I. DIRECT (Con't) 10F2 5.8 miles SSW oi or1g1n Collected quarterly Gamma dose quarterly lOGl # 11.6 miles SSW of origin llSl 0.07 miles SW of origin 11E2
    • 5.0 miles SW of origin
    . llFl 5.2 miles SW of origin 12El 4.4 miles WSW of origin 12Fl 9.4 miles WSW 6r origin 13El . 4.2 miles NE of origin 13F.4 9.8 miles W of origin 13F2 6.5 miles W of origin 13F3 9.3 miles W of origin 1401 3.4 miles WNW of origin 14F2 6.6 miles WNW of origin 15F3 5.4 miles NW of origin 16El 4.1 miles NNW of origin 16F2 8.1 miles NNW-of origin 16Gl # 14.8 miles NNW of origin
    1. Control Station (in addition to controls listed, two additional program controls are used for internal studies and are referred to as SITE-CAL and SITE-ZERO).
    M P85 183/15 4-dh Rev. 3 7/30/87
    • TABLE E-1 '
    ODCM - SALEM GENERATING STATION Page 3 of 7 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE TYPE AND FREQUENCY PATHWAY STATION CODE LOCATION COLLECTION METHOD OF ANALYSES II. AIRBORNE (a) P 2S2 0.4 miles NNE of origin Continous low volume air ~ross beta analysis A sampler. Sample collected on each weekly R every week along with filter sample. Gamma T change. spectrometry shall I 2F2 8.7 miles NNE of origin be performed if c gross beta exceeds L ten times the A yearly mean of T control station E value. s 3H3 # 110 miles NE of origin Gross beta analysis done > 24hr. after sampling to allow for Radon and Theron daughter decay 1001 3.9 miles SSW of origin 16El 4.1 miles NNW of origin Gamma isotopic analysis on quarterly composite
    1. Control Station M P85 183/15 4-dh Rev. 3 7 /30/87
    • **TABLE E-1 Page 4 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE TYPE AND FREQUENCY PATHWAY STATION CODE LOCATION COLLECTION METHOD OF ANALYSES II. AIRBORNE (Con't)
    ( b) I 2S2 0.4 miles NNE of origin A TEDA impregnated charcoal Iodine 131 analyses 0 flow-through cartridge i~ are performed on D connected to.air particulate each weekly sample. I air sampler and is collected N weekly at filter change. E 2F2 8.7 miles NNE of origin 3H3 # 110 miles NE of origin 16El 4.1 miles NNW of origin 1001 3.9 miles SSW of origin
    1. Control Station M P85 183/15 4-dh Rev. 3 7/30/87
    • TABLE E-1 Page 5 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY STATION CODE LOCATION COLLECTION METHOD ANALYSES III. WATER (a) S 7El 1 mile W of Mad Horse Sample to be collected Gamma isotopic u Creek; 4.5 miles SE of monthly providing winter analysis monthly R origin icing conditions allow F sample collection H-3 on quarterly A composite c 12Cl # West bank opposite E Artificial Island; 2.5 miles WSW of origin 16Fl C&D canal; 6.9 miles NNW of origin
    ( b) G 2S3 Fresh water holding tank; R 700 feet NNW of origin 0 5Dl Local farm; 3.5 miles E Collected monthly Gamma isotopic u of origin monthly N 3El Local farm; 4.1 miles Tritium analysis D NE of origin monthly , ( c) s 7El 1 miles W of Mad Horse A sediment sample is Gamma isotopic E Creek 4.5 miles SE of taken semi~annually anaylsis - semi-* D origin annually I* 12Cl # West bank opposite M Artificial Island; 2.5 E miles WSW of origin N 16Fl C&D Canal; 6.9 miles NNW T of origin
    1. Control Station M P85 183/15 4-d.h Rev. 3 7 /30/87
    TABLE E-1 Page 6 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY STATION CODE LOCATION COLLECTION METHOD ANALYSES IV. INGESTION {a) M 2F7 5.7 miles NNE of origin Sample of fresh milk is
    • Gamma isotopic and I collected for each farm I-131 analyses on L semimonthly when cows each sample on K are on pasture, monthly collection at other times.
    3Gl # 16.6 miles NE of origin 5F2 7.0 miles E of origin 13E3 4.9 miles w of origin 14Fl 5.5 miles WNW of origin { b) F llAl Outfall area; approx. Two batch samples of fish Gamma isotopic I 650 feet SW of origin are sealed in plastic bag analysis of edible s or jar and frozen portion on H semiannually or when in collection in season l2Cl # West bank opposite Artificial Island;* 2.5 miles WSW of origin
    1. Control Station M P85 183/15 4-dh Rev. 3 7 /30/'d7
    • *TABLE E-1 Page 7 of 7 ODCM - SALEM GENERATING STATION RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM PATHWAY STATION CODE LOCATION COLLECTION METHOD ANALYSES IV. INGESTION (Cont'd)
    ( c) I llAl Outfall area; approx. Two batch samples of crab Gamma Isotopic N 650 feet SW of origin are sealed in a plastic analysis of v bag oi jar and frozen edible portion E semiannually or when in on collection R season. T 12Cl # West bank opposite E Artificial Island; B 2.5 miles WSW of origin R A T E s
    1. Control Station M P85 183/15 4-dh Rev. 3 7/30/87
    FIGURE E-1 OFFSITE SAMPLING LOCATIONS ARTIFICIAL ISLAND e'..........***** , \
    • *e i!
    \. Rev. 3 7/30/87 FIGURE E-2 ONSITE SAMPLING ARTIFICIAL LOCATIONS
    • ISLAND
    \ \
    • Rev. 3 7/30/87
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